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

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At least 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
NASA NTRS2018

Electrically Heated Testing of the Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Experiment Using a Depleted Uranium Core

The Kilopower project aims to develop and demonstrate scalable fission-based power technology for systems capable of delivering 1 to 10 kW of electric power with a specific power ranging from 2.5 to 6.5 W/kg. This technology could enable high-power science missions or could be used to provide surface power for manned missions to the Moon or Mars. NASA has partnered with the U.S. Department of Energy's National Nuclear Security Administration, Los Alamos National Laboratory, Nevada National Security Site (NNSS), and Y−12 National Security Complex to develop and test a prototypic reactor and power system using existing facilities and infrastructure. This technology demonstration, referred to as the "Kilowatt Reactor Using Stirling TechnologY (KRUSTY)," will undergo nuclear ground testing by the end of calendar year (CY) 2017 at the NNSS. The 1-kWe variation of the Kilopower system was chosen for the KRUSTY demonstration. The concept for the 1-kWe flight system consists of a 4 kWt highly enriched uranium-molybdenum reactor operating at 800 degC coupled to sodium heat pipes. The heat pipes deliver heat to the hot ends of eight 125-W Stirling convertors producing a net electrical output of 1 kW. Waste heat is rejected using titanium-water heat pipes coupled to carbon composite radiator panels. The KRUSTY test, based on this design, uses a prototypic highly enriched uranium-molybdenum core coupled to prototypic sodium heat pipes. The heat pipes transfer heat to two Advanced Stirling Convertors (ASC−E2s) and six thermal simulators, which simulate the thermal draw of full-scale power conversion units. Thermal simulators and Stirling engines are gas cooled. The most recent project milestone was the completion of nonnuclear system-level testing using an electrically heated depleted uranium (DU) (nonfissioning) reactor core simulator at the NASA Glenn Research Center. System-level testing has validated performance predictions and has demonstrated system-level operation and control in a test configuration that replicates the one to be used at the Device Assembly Facility (DAF) at the NNSS. Fabrication, assembly, and testing of the DU core has allowed for higher fidelity system-level testing at Glenn, and has validated the fabrication methods to be used on the highly enriched uranium core that will supply heat for the DAF KRUSTY demonstration.

Briggs, Maxwell H.↗
DOE OSTI2021

Update on the High Precision Titration Method for Uranium Assay Supported by NBL Program Office

The NBL Program Office (NBL PO) has coordinated with Oak Ridge National Laboratory (ORNL) to implement the High Precision Titration (HPT) method for uranium assay measurements. The measurement method has been successfully developed and qualified, therefore ORNL has progressed to performing analyzes critical to the mission of the NBL PO. The HPT method is vital to producing the next generation of certified reference materials for uranium assay and isotopic abundance. The work described here focused on establishing traceability of the method to the NIST SRM 136 potassium dichromate series, and focused on investigating small differences between SRM 136e and 136f. The NBL PO evaluated the history of all of its primary uranium reference materials, including CRM 112A natural uranium metal, and worked with NIST in investigating the small differences in the dichromate SRM’s. ORNL performed comparative experiments on SRM 136e and 136f using CRM 112A. The results of the experiment will be presented here, along with NBL PO’s plans to re-evaluate the CRM 112A certified uncertainty to comply with JCGM 100, “Guide to the expression of uncertainty in measurement.”

Rogers, Kayron↗
DOE OSTI2024

Commercialization of High-Density High Assay Low Enriched Uranium Fuel Systems

The Office of Reactor Conversion and Uranium Supply (NA 231) at the National Nuclear Security Administration leads the conversion effort for the United States High Performance Research Reactors (USHPRR). These reactors are the final civilian reactors in the US to transition from High Enriched Uranium (HEU) to high assay low enriched uranium (HALEU). Each of these reactors represents unique capabilities and no currently available fuel system meets their needs for conversion. The Fuel Fabrication (FF) Pillar of the USHPRR project is responsible for the fabrication of experimental elements, conversion elements, and establishing a commercial economical production capability. FF is also responsible to share with the domestic and international community the theoretical knowledge gained. Other pillars within the USHPRR project provide the experimental and conversion fuel designs, assist the reactors with licensing activities, and ensure the entire fuel cycle is evaluated. Over the last decade, FF has worked with the production partners at Y-12 National Security Complex (Y-12) and BWXT Nuclear Operations Group, Research and Test Reactors (BWXT). Y-12 has begun processing the alloy feedstock for the conversion elements with a qualified process. BWXT has started the final fabrication of the experimental elements. Once the experimental elements are complete, BWXT will begin conversion element fabrication. The FF Pillar resides at Pacific Northwest National Laboratory (PNNL) and uses PNNL, universities, commercial vendors, and the DOE national laboratory system to evaluate process development activities to improve the process steps. FF supports the fabrication of two high density fuel systems, monolithic U-10Mo (Figure 1) and Uranium Silicide (Figure 2). The U-10Mo fuel system is further along the development process. FF assists in long term planning with the production partners. This includes ramping production of the elements from experimental quantities to annual steady state needs. As part of the ramp up, opportunities to improve yield and product quality are identified to ensure the fuel systems are cost effective.

Catalan, Michael A. [BATTELLE (PACIFIC NW LAB)]↗
DOE OSTI2026

Impact of Molten Gallium on the Microstructure and Corrosion Behavior of Aluminum and Uranium-Aluminum Alloys for Used Nuclear Fuel Reprocessing

Test reactors around the world utilize highly enriched uranium fuel to achieve high neutron fluxes for materials testing. Once spent, the remaining uranium is a valuable resource for subsequent fuel fabrication. However, some of these test reactor cores consist of curved plate-type fuel elements, fabricated using aluminum alloy 6061 (AA6061) cladding to encapsulate a uranium-aluminum alloy (UAlx) fuel matrix. These assemblies require non-standard reprocessing approaches for uranium recovery, as aluminum dissolves readily in acidic solutions, generating large volumes of waste and complicating downstream chemical separations. In this work, we investigate a novel chemical decladding strategy based on the interaction between AA6061/UAlx and molten gallium (Ga). Ga is known to induce severe degradation of aluminum metal through liquid metal embrittlement (LME), even at relatively low Ga concentrations. By penetrating the aluminum crystal lattice, Ga disrupts grain cohesion and facilitates fracture or dissolution of the aluminum matrix. Thermodynamic analysis of the Al–Ga binary phase diagram suggests that Ga may offer a viable pathway to selectively weaken or dissolve the AA6061 cladding, and potentially the aluminum component of the UAlx fuel matrix within. To this end, parametric experiments were performed at 50 °C and 100 °C across a range of Al–Ga atomic fractions. At lower Al fractions, the AA6061 was completely molten after 2 hours of exposure to the Ga metal. In contrast, samples with higher Al fractions (0.9 Al, 0.1 Ga) contained residual solids after 2 hours, which were characterized by microstructural examination using electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). These Al-Ga compositions were also evaluated using FactSage thermodynamic modeling to further elucidate the relationship between phase diagram behavior and LME.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Nuclear forensic analysis on uranium ore concentrates: A multi-laboratory intercomparison study

A collaboration between the United States and Japan performed nuclear forensic analyses on a set of 7 uranium ore concentrate (UOC) samples with known origins to improve methodologies and best practices in the field. Overall, good agreement between the 4 participating laboratories was achieved for uranium assay, uranium isotope composition, strontium isotope composition, and trace element analysis. In conclusion, the combination of these four signatures is sufficient to distinguish the different origins among these studied UOCs.

and nuclear chemistry↗
DOE OSTI2023

Computational studies of impurity migration during induction stirring of molten uranium

Understanding and controlling impurity behavior in actinide metal casting processes are foundational for efficient part production yet remain major challenges for researchers and industry. To help provide insight regarding impurity distribution during actinide metal casting, we have developed computational fluid dynamics (CFD) models for a laboratory-scale system using commercial and open-source code. Multiple simulation frameworks allow for improved confidence in the resulting outputs, while taking advantage of the maturity and convenience offered by commercial platforms and simultaneously maintaining the transparency and flexibility often provided by open-source software. Here, we describe multiple experiment-informed models designed to simulate a specific laboratory system in which uranium melt, containing a known starting concentration of carbon impurity, is electromagnetically stirred in an induction furnace. Here, the goal of the simulations is to predict the motion of uranium carbide microparticles in the velocity field of the melt. Prior to simulating the uranium-carbon system, numerical models were validated using a previously published nonradioactive experimental system. Effects of the size and shape of impurity particles in the models were investigated and agree with experimental findings. Simulation of smaller particles (< 50 µm) shows more homogenous distribution throughout the stirred melt. With increased particle size (100 µm), the body forces, which include the buoyancy force, dominate over the drag force, causing larger particles to move toward the crucible walls and upward in the system.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Thermal impacts on nondestructive analysis measurements of uranium hexafluoride

A series of nondestructive assay measurements were performed before, during, and after localized heating of an industrial standard 12B cylinder of uranium hexafluoride to replicate dose variations previously observed on commercial 30B cylinders. While cylinders represent a closed system, enrichment measurements may be impacted by altered uranium and daughter product distributions due to sublimation. A series of measurements were performed to examine localized heating impacts on dose rates, ratios of uranium and daughter product distributions, and measured 235 U enrichment as calculated by two different approaches. The less complex enrichment meter approaches were found to be more robust and less prone to deviation than isotope ratio approaches such as FRAM, but the impact can be minimized by using observable dose changes to avoid impacted areas.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Separation of protactinium from uranium-niobium alloys for 231 Pa– 235 U radiochronometry in nuclear forensic investigations

The isolation and purification of protactinium from uranium materials is essential for 231 Pa– 235 U radiochronometry, but separating Pa from uranium-niobium alloys, a common material in the nuclear fuel cycle, is challenging due to the chemical similarity of Pa and Nb. Here, in this work, we present three resin chromatography separation techniques for isolating Pa from U and Nb which were independently developed by three different laboratories through ad hoc adaptations of standard operating procedures. Our results underscore the need for and value of purification methods suitable for a diversity of uranium-based materials to ensure the operational readiness of nuclear forensics laboratories.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

Radiochronometric discordance in cast uranium metal: a multi-laboratory intercomparison exercise

Here, the model age of a nuclear material is crucial in nuclear forensic analysis. Uranium metals with complex production histories often exhibit discordant model ages from the 230 Th– 234 U and 231 Pa– 235 U chronometers. Recent studies involving targeted uranium metal castings have enhanced our understanding of decay product behavior during casting, aiding nuclear forensic interpretation. Building on this prior work, forensics laboratories at Atomic Weapons Establishment (AWE), Lawrence Livermore National Laboratory (LLNL), and Los Alamos National Laboratory (LANL) conducted an interlaboratory comparison to investigate spatial heterogeneity in uranium metal cast under controlled conditions. Each laboratory measured samples of a mixed feedstock and its corresponding cast product. This work furthers our understanding of discordant model ages and the use of discordance as a signature to enhance confidence in interpretations of radiochronometric data for nuclear forensics.

230Th/234U↗
DOE OSTI2024

Evolution of dislocation loops in irradiated α-Uranium: An atomistically-informed cluster dynamics investigation

An atomistically-informed mean field cluster dynamics model has been developed to investigate the nucleation and growth of prismatic loops in irradiated a-Uranium. TEM analysis of neutron irradiated a-Uranium shows the evolution of self-interstitial atom and vacancy loops on (010) and (100) crystallographic planes, respectively, resulting in an anisotropic lattice swelling of its face-centered orthorhombic crystal. To provide model parameters, the crystallography of loops and the binding energy of point defects to these loops were studied using an angular dependent EAM potential and classical molecular dynamics (MD) simulations. Furthermore, using the bond-boost hyperdynamics method, the anisotropic diffusion of interstitials and vacancies in a-Uranium was investigated. Here, the mechanisms of point defect diffusion and the associated migration energies were reported and compared with previous DFT studies. The energetics and kinetic quantities mentioned above were used in the cluster dynamics model to predict the cluster density as a function of dose rate, dose and temperature and the results were compared to the reported neutron irradiation experiments. The model predictions reveal an accumulation of small sized vacancy loops along with a population of large and growing self-interstitial loops, which closely corresponds to the TEM observations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

CALPHAD modeling of uranium nitride (UN) fabrication routes enabled by first-principles calculations

The thermochemical details of fabricating uranium nitride (UN) by ammonolysis of uranium tetraflouride (UF 4 ) were determined using density functional theory (DFT) and CALculation of PHAse Diagrams (CALPHAD) computational methods. The thermochemical data of all binary, ternary, and quaternary U-H-N-F phases were computed using DFT, and the data for the phases that have not been measured experimentally, including UN 2 and NH 4 F(g), were combined with existing experimentally-determined data for CALPHAD modeling. The DFT data were benchmarked using experimental Gibbs energy of reaction and experimental thermochemical data for individual species. Phase diagrams relevant to the ammonolysis reaction are depicted, showing regions of stability for solid U-N, U-F and U-N-F phases. An unidentified phase produced in a previous experiment was identified as UN 0.95 F 1.2 (UNF) by comparing its X-ray diffraction spectrum to the experimental spectrum, and its formation during the fabrication of UN from UF 4 is supported by the simulated phase diagram. Here, it is calculated that UN 2 can be produced by the ammonolysis of UF 4 , but requires elevated temperatures, high NH 3 (g) partial pressure, and large amounts of flowing NH 3 (g) to avoid solid flu oride impurities in the uranium nitride. Likewise, U 2 N 3 can be produced instead at temperatures greater than 980 K. The use of silane (SiH 4 ) gas was investigated as a potential additive in the ammonolysis fabrication route to speed removal of fluorine. The addition of SiH 4 (g) offers little advantage to the removal of fluorine, and adds the complication of Si 3 N 4 formation. The use of DFT to fill in missing data to perform CALPHAD calculations demonstrated here allows for the determination of more comprehensive and trustworthy phase diagrams than the use of existing experimental data alone.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2026

Preparation of a uranium monocarbide anode and electrochemical characterization in molten LiCl-KCl-UCl 3

Porous uranium carbide (UC) pellets possessing moderate electrical conductivity were synthesized by reaction of UO 2 with graphite at temperatures up to 1550°C under rough vacuum. Conversions as high as 98% were achieved at soak times of 2-4 hours. The electrochemistry of the UC pellets in molten LiCl-KCl-6.5 wt% UCl 3 was explored using a variety of techniques including DC polarization methods, cyclic voltammetry, chronopotentiometry and bulk electrolysis. Here, the electrode reaction for anodic dissolution was found to be kinetically controlled by dissociation of UC to a transition state complex that was hypothesized to consist of a uranium atom partially complexed by chloride ions. Precise measurements of current efficiencies using chronopotentiometry indicated upper limits of 90.9 ± 3.4% and 98.3 +1.7/-3.7% for anode and cathode, respectively, when operating at anodic overpotentials near +300 mV. Bulk electrolysis of a UC pellet performed by passing 98% of the theoretical charge resulted in nearly complete recovery of its uranium content as highly pure metal at the cathode.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Faraday detector uranium isotope ratio measurement: Insights from solution- and laser ablation-based sampling methodologies on the Neoma MC-ICP-MS

In this study we utilize an all-faraday cup detector configuration on the Neoma MC-ICP-MS to perform uranium isotope ratio determinations from certified reference materials and a nuclear fuel precursor material via solution and laser ablation based sampling methodologies. The goal in performing these measurements is primarily to demonstrate that the all-faraday cup method can produce highly precise and accurate isotope ratios even at relatively low signal intensity levels while also developing a preliminary understanding of how the complex interplay between different amplifier resistor levels (e.g. 10 11 vs 10 13 Ω) and integration times impacts the precision and accuracy of different types of measurements. Our results show that the faraday cups with 1013 Ω resistors in their amplifier feedback loops can be used to produce accurate and relatively precise isotope ratios for ion beams down to ~5 K cps. However, at the lower end of the signal intensities observed in this study, longer integration times are necessary to mitigate the impact of ion beam instability that can occur during laser ablation based sampling. Despite this potential limitation, the all-faraday cup method is highly versatile as exemplified by the analysis of a collection of uranium reference materials with highly variable uranium isotopic compositions.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Best practices for predictions of radionuclide activity concentrations and total absorbed dose rates to freshwater organisms exposed to uranium mining/milling

Predictions of radionuclide dose rates to freshwater organisms can be used to evaluate the radiological environmental impacts of releases from uranium mining and milling projects. These predictions help inform decisions on the implementation of mitigation measures. The objective of this study was to identify how dose rate modelling could be improved to reduce uncertainty in predictions to non-human biota. For this purpose, we modelled the activity concentrations of 210 Pb, 210 Po, 226 Ra, 230 Th, and 238 U downstream of uranium mines and mills in northern Saskatchewan, Canada, together with associated weighted absorbed dose rates for a freshwater food chain using measured activity concentrations in water and sediments. Differences in predictions of radionuclide activity concentrations occurred mainly from the different default partition coefficient and concentration ratio values from one model to another and including all or only some 238 U decay daughters in the dose rate assessments. Consequently, we recommend a standardized best-practice approach to calculate weighted absorbed dose rates to freshwater biota whether a facility is at the planning, operating or decommissioned stage. At the initial planning stage, the best-practice approach recommend using conservative site-specific baseline activity concentrations in water, sediments and organisms and predict conservative incremental activity concentrations in these media by selecting concentration ratios based on species similarity and similar water quality conditions to reduce the uncertainty in dose rate calculations. At the operating and decommissioned stages, the best-practice approach recommends relying on measured activity concentrations in water, sediment, fish tissue and wholebody of small organisms to further reduce uncertainty in dose rate estimates. This approach would allow for more realistic but still conservative dose assessments when evaluating impacts from uranium mining projects and making decision on adequate controls of releases.

63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗
DOE OSTI2021

Uranium nitride (UN) pellets with controllable microstructure and phase – fabrication by spark plasma sintering and their thermal-mechanical and oxidation properties

Dense uranium mononitride (UN) pellets with controlled microstructures and tailored grain size from large-grained to a few microns are synthesized by spark plasma sintering (SPS) combined with high energy ball milling. The impacts of the sintering conditions on fuel microstructure, grain size, physical density, and phase behavior are systematically investigated, and the thermal-mechanical properties and oxidation behavior of the SPS densified UN pellets are characterized. Higher sintering temperatures and longer ball milling durations and thus finer starting UN powders promote sintering and densification, and dense UN pellets above 95% theoretical density can be achieved by SPS at 1873 K for 10 min. UN phase purity is maintained in the SPS-densified pellets sintered at a lower temperature and short duration. A phase heterogeneity with secondary UO 2 or uranium sesquinitride (U 2 N 3 ) occurs for the UN pellets sintered at higher temperatures using finer UN powders. The hardness and fracture toughness of the SPS-densified UN pellets increase with smaller grain sizes and higher densities to 7.9 GPa and 3.5 MPa m 1/2 , respectively. Furthermore, both small (1–2 μm) and large grain-sized (30–50 µm) UN pellets exhibit good thermal conductivity. Dynamic oxidation testing by a thermogravimetric analyzer in air shows that the onset temperature for oxidation varies with microstructure and phase heterogeneity of the SPS densified UN pellets. Particularly, the smaller-grained (micron-sized) UN pellets containing uranium oxides and U 2 N 3 display lower weight gain and significantly-reduced oxidation kinetics, and full oxidation completes at a temperature above 1173 K when tested with a ramp rate of 10 K/min.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Advancing Elemental and Isotopic Analysis of Uranium Mineral Inclusions: Rapid Screening via Laser-Induced Breakdown Spectroscopy and High-Resolution Laser Ablation-ICP-MS Mapping

The work presented herein employs two laser-based analytical techniques (laser-induced breakdown spectroscopy (LIBS) and laser ablation – inductively coupled plasma – mass spectrometry (LA-ICP-MS)) to spatially determine the elemental and isotopic composition of uranium bearing minerals. Uniquely, this work leverages the high-speed applicability of LIBS to “screen” the sample(s) for their elemental constituents. After determining the location of the uranium inclusions (via LIBS), high-resolution LA-ICP-MS was employed to further characterize the inclusions. The high-resolution (sub-µm) capabilities of LA-ICP-MS were able to extract important information from the uranium minerals including discerning its chemical form (e.g., finchite from carnotite, Sr- and K-bearing uranyl vanadates, respectively) as well as their 235 U/ 238 U isotopic composition. This approach, LIBS followed by LA-ICP-MS, significantly reduces the analysis time (~95 %) in comparison to employing a LA-ICP-MS only approach. Furthermore, this work presented a novel approach to analyzing inclusions via a particle/inclusion analysis tool which is commercially available within the iolite 4 software. This tool allowed for a more accurate characterization of the isotopic distribution of the inclusions, as well as allowing for rapid sizing of the inclusions. Finally, this analytical approach could readily be applied to other sample types in which the target species (e.g., µm-sized inclusions) are embedded in complex matrices (e.g., cm-sized samples).

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Evaluation of flow-induced plate deflection for University of Missouri research reactor low-enriched uranium fuel element

The University of Missouri Research Reactor (MURR), located on the campus of the University of Missouri in Columbia, Missouri, is one of the six United States (U.S.) High Performance Research Reactors (USHPRR), including one critical facility, that are actively collaborating with the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA) Office of Material Management and Minimization (M3) Reactor Conversion Program to convert from highly enriched uranium (HEU, ≥20 wt% U-235) fuel to low-enriched uranium (LEU, <20 wt% U-235) fuel. A new type of very high-density LEU fuel based on a monolithic alloy of uranium and 10 wt% molybdenum (U-10Mo) is expected to allow conversion of some USHPRR, including MURR. In the design of its fuel elements, MURR is using thin parallel curved fuel plates separated by coolant channels. In this work, fluid-structure interaction (FSI) analysis of the MURR LEU fuel element is performed at the element level (as compared to the plate level analysis), which models all components of the LEU fuel element, including fuel plates and the supporting structures. Therefore, the effect of supporting structures on the flow distribution within the element and the fuel plate deflection are evaluated. In addition to the element nominal flow rate and dimensions, the tolerances in the geometry of the coolant channel and plate thickness, the effect of a comb on plate deflection, and the uncertainty of the flow rate per element are evaluated. For the LEU fuel plates, which are thinner than the current HEU plates, the predicted plate deflection is found to be small compared to the fabrication and assembly tolerances. Thus, the FSI-induced deflections are not expected to noticeably reduce the coolant flow rate or predicted safety margins in the limiting channels for the MURR LEU fuel element. In addition to the simulation work, a hydraulic performance test of the MURR LEU fuel element is currently being planned to support conversion to the use of LEU fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Preliminary conceptual design of fast neutron spectrum nuclear thermal rocket cores using monolithic uranium nitride fuel

This paper presents a few nuclear thermal rocket (NTR) reactor core preliminary conceptual designs based on the use of monolithic uranium nitride (UN) fuel plates or pins clad with tungsten (W), alloyed or unalloyed as needed. High-assay low-enriched uranium (HALEU) as well as high-enriched uranium (HEU) are considered. Nominal core thermal powers are between 300 and 1100 MW corresponding to thrusts between 66 kN and 240 kN. The estimated thrust-to-reactor-weight ratios (T/WRx) achievable with the HEU plate and pin configurations are between 3.8 and 10.9. The T/WRx achievable with the HALEU plate and pin configurations are lower, 1.9–4.0, than those achievable with HEU. It is noteworthy that even a relatively modest increase in fuel enrichment above the HALEU limit enables a significant increase in T/WRx. It must be emphasized that these results do not yet account for uncertainties and that the appropriateness of the assumed fuel thermal design limits for normal operation (i.e., peak centerline temperature of 3100 K) will need to be confirmed. Finally, just like for the other UN-based fuels currently considered for NTR applications (i.e., CERMET and CERCER), the proposed UN fuel systems need significant testing in prototypic conditions to confirm the appropriate performance for safe NTR applications. However, the monolithic fuel systems may have, overall, a higher technology readiness level (TRL), and, consequently, shorter fuel development and validation time may result.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗