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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 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
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 OSTI2020

Initial development of an RIA envelope for dispersed nuclear fuel

A reactivity insertion accident (RIA) is a design basis accident in which reactivity is rapidly injected as a result of a control rod ejection or blade drop scenario. The resultant power increase can result in fuel rod failure and subsequent release of radioactive material into the reactor’s primary system. For light water reactors, failure under RIAs is typically a result of fuel melt, pellet-cladding interaction, or rod over-pressurization. Fuel melt occurs when the fuel system is unable to transport heat out of the fuel system. Pellet-cladding interaction occurs when an aggressive fuel pellet expansion pushes on an embrittled cladding beyond its strain limit. Rod over-pressurization occurs when a fuel rod exceeds the critical heat flux and departs from nucleate boiling, causing the rod internal pressure to rapidly increase, exceeding the systems pressure, and balloon until it bursts. To prevent failure, regulators have developed safety requirements that limit the injected enthalpy based on the state of the fuel system. However, dispersed fuel could enable the removal or relaxation of regulator-imposed safety criteria. Dispersed fuel embeds fuel particles in a highly conductive metal. Dispersed particles can have a particle radius of 1 µm to greater than 100 µm. Traditional fuel systems are plagued by poor thermal conductivity, especially at higher burnups, thus reducing the ability of the fuel system to respond to an RIA. However, dispersed fuel could improve the ability of the fuel system to transport injected energy away from the fissile material and into the coolant more efficiently. Furthermore, dispersing the fuel particles mitigates hard contact that typically occurs in the traditional zirconium-uranium dioxide fuel system, thus removing pellet-cladding mechanical interaction as a potential failure mechanism. This paper evaluates fuel particles dispersed in a metal matrix using the BISON fuel performance code to develop an initial failure threshold based on melt temperature of the fuel particle and/or cladding material as a function of fission density for beginning-of-life and end-of-life conditions. BISON results show that (1) reducing the size of the particles allows fission density to increase as a function of time, (2) particle-to-particle proximity must be considered to evaluate the limiting conditions, and (3) improving the fuel particle thermal conductivity improves thermal performance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Status of the Fuel Fragment Dispersal Studies at Oak Ridge National Laboratory

During a loss-of-coolant accident, a nuclear fuel rods undergo balloon and burst. High burnup fuel is known to fragment and pulverize leading to the potential for the fuel to disperse through the burst opening either during the initial burst or following the burst during the reflood event. This milestone summarizes the design and assembly of a system to study dispersal in a postburst event. Surrogate fuel material (HfO2, tungsten, and yttria-stabilized zirconia [YSZ]) dispersing from as-burst loss-of-coolant accident (LOCA) test specimens was extensively investigated, along with dispersal of actual nuclear fuel following a LOCA Severe Accident Test Station (SATS) test. Surrogate material tests spanned a variety of burst geometries, particle types, particle size distributions, and rod oscillation loadings (frequency and amplitude combinations).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Status of the Fuel Fragment Dispersal Studies at Oak Ridge National Laboratory

During a loss-of-coolant accident, a nuclear fuel rods undergo balloon and burst. High burnup fuel is known to fragment and pulverize leading to the potential for the fuel to disperse through the burst opening either during the initial burst or following the burst during the reflood event. This milestone summarizes the design and assembly of a system to study dispersal in a postburst event. Surrogate fuel material (HfO 2 , tungsten, and yttria-stabilized zirconia [YSZ]) dispersing from as-burst loss-of-coolant accident (LOCA) test specimens was extensively investigated, along with dispersal of actual nuclear fuel following a LOCA Severe Accident Test Station (SATS) test.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Temperature Effects on Interdiffusion of Al and U-Mo under Irradiation

A high-energy Xe ion irradiation experiment was conducted to investigate the temperature dependence of interdiffusion in bilayer Al-UMo samples under irradiation. The amount of interdiffusion achieved at a fixed dose with the increase of temperature showed a clear transition at 175 degrees C (with an estimated error in the range of ± 10 degrees C) from temperature-independent to temperature-dependent behavior. The activation energy derived from the curve of interdiffusion quantity vs. irradiation temperature is 0.77 ± 0.16 eV. This information has been utilized to understand the temperature effect on the interdiffusion process that occurred at the interfaces of U-Mo particles and the Al matrix in U-Mo/Al dispersion fuels, whose magnitude significantly impacts the fuel's performance. Although this temperature effect was deemed important, it cannot be examined directly using in-pile irradiation data, as fuel temperatures cannot be measured in reactor irradiation and are highly correlated with fission rate and thermal conductivity evolution. To connect the knowledge accumulated from ion irradiation with in-pile irradiation data, simulation of a full-sized U-Mo/Al dispersion fuel plate irradiated in the FUTURE test in the BR2 reactor was performed with the Dispersion Analysis Research Tool (DART), a dispersion fuel performance code. DART is equipped with an interaction or interdiffusion layer (IL) growth correlation formulated to describe the temperature dependence of ion mixing results. The agreement between calculated and measured fuel meat constituent volume fractions and swelling data demonstrated that the temperature effect on in-pile Al-UMo interdiffusion is well captured with the correlation. In this case, the fitted activation energy is 0.70 eV. Considering the uncertainties associated with the ion irradiation data, the activation energy obtained from in-pile data fitting is in accord with that from ion irradiation results.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Forming Complex Nuclear Fuel Shapes in High-Loaded Silicide Surrogates

This work provides proof of the concept that high silicide loading nuclear fuel meat surrogates with complicated geometries can be produced with uniform density through an application of cold isostatic pressing (CIP). Dispersion fuels with high volumetric loading of U_3 Si_2 have challenges in fabrication. Fabrication involves a series of processes including powder compaction, rolling pack assembly, and roll forming. PNNL has undertaken extensive experimental work using MoSi2 and WSi2 as surrogates for U3Si2 to explore the feasibility of reducing or eliminating the issues through the application of CIP in the powder pressing step prior to rolling pack assembly. The composites were prepared at >40 vol% silicide loading, which was representative of 4.8 gU/cm^3 and formed via CIP at 50 KSI (Kilopound per Square Inch) pressure. The CIP mold design was taken through a design process which aimed to reduce defects and increase precision. The application of CIP here provides a broad strategy for producing highly loaded dispersion fuels with complex geometries and uniform density.

Clelland, Dustin T.↗
NASA NTRS1974

Theoretical nitric oxide production incidental to autoignition and combustion of several fuels homogeneously dispersed in air under some typical hypersonic flight conditions

A reaction package of 100 chemical reactions and attendant reaction rate constants defined for the autoignition and combustion of four carbonaceous fuels, CH4, CH3OH, C2H6, and C2H5OH. Definition of the package was made primarily by means of comparison between trial calculations and experimental data for the autoignition of CH4. Autoignition and combustion of each of these four fuels was calculated under three sets of conditions realistic for hypersonic flight applications, for comparison to hydrogen fuel, particularly with respect to formation of nitric oxide. Results show that, for all of the fuels including hydrogen, if NO production is a significant problem, compromise must be made between approaching equilibrium heat release and approaching equilibrium NO concentration.

Bahn, G. S.↗
DOE OSTI2020

Thermal conductivity measurement of the interaction layer between UMo and Al produced by high-energy heavy ion irradiation

In this work, we report the first direct thermal conductivity measurement results for Al-UMo interaction layer (IL), which is typically observed in UMo/Al dispersion fuel plates under irradiation. The investigated IL was formed by irradiating Al coated UMo substrate using 80 MeV iodine ions at 180 degrees C up to 3.03 x 10 17 ions/cm 2 fluence. Microstructural characterization indicated that the induced IL is amorphous with an approximately (U 0.8 ,Mo 0.2 )Al 5.3 stoichiometry, which is similar to that formed under in-pile irradiation. Focused ion beam (FIB) was used to prepare nine specimens of various lengths from the IL that could be suspended across a microfabricated device for thermal conductivity measurement. The measured thermal conductivity values of the IL were significantly lower than the values for both the original UMo fuel and the Al. The successful measurement of the Al-UMo IL provides valuable information for the development and qualification of UMo/Al dispersion fuels for research and test reactor conversion applications, and further demonstrates the promising capabilities of utilizing the suspended bridge method in nuclear fuel research.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Mapping the Microstructure During the Historic U-Mo Monolithic Fuel Foil Fabrication Process

Since 2004, there has been extensive effort towards the development of a uranium molybdenum monolithic fuel system to convert high performance research and test reactors. The RERTR-6 experiment was the first to attempt a monolithic fuel instead of a dispersed fuel form. The fabrication methods evolved overtime and provided the basis for current fabrication methods. The various steps in the process inevitably tailor the fuel alloy microstructure which is known to influence irradiation behavior. This document aims to present and discuss the fabrication evolution that transpired through a recounting of historical data from the various fabrication campaigns. By overlaying this data with basic science studies on the U-Mo that explored transformation kinetics, it is possible to estimate a measure of the impact heat treatments have on the final as-fabricated microstructure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Effect of spacer grids on high-burnup fuel fragmentation, relocation, and dispersal

Increasing the fuel burnup limit in light-water reactors to improve fuel cycle economics requires a strong technical foundation. Experimental observations from the Halden and Studsvik programs have revealed severe fuel fragmentation during loss-of-coolant accident (LOCA) conditions, highlighting the need for additional technical evaluation. Consequently, further LOCA test data are needed to complement existing findings and improve the understanding of fuel fragmentation, relocation, and dispersal (FFRD) behavior. Oak Ridge National Laboratory’s Severe Accident Test Station has played a significant role in advancing the understanding of high-burnup fuel fragmentation, relocation, and dispersal phenomena. One remaining gap in the available experimental database is the effect of fuel assembly structural features on cladding deformation behavior during a LOCA, and more specifically, their impact on the fuel’s ability to fragment, relocate, and disperse. Recent analyses using the BISON fuel performance code suggest that cladding deformation near grid spacers will remain below the 3% threshold that has been reported in the NRC Research Information Letter, indicating that the cladding could remain mechanically constrained during the LOCA event. This paper builds upon the BISON analyses to design and conduct a series of out-of-cell tests aimed at further evaluating cladding deformation in and around grid spacers. In addition, these tests were used to assess local cladding temperature conditions and compare them against analytical predictions in order to better replicate expected in-reactor behavior. Finally, an in-cell high-burnup LOCA test was designed and performed to evaluate the effects of a grid spacer, or cladding restraint, on fuel fragmentation, relocation, and dispersal susceptibility. The high-burnup test results differed from those of historical LOCA experiments, with a recorded rupture temperature of 861°C. Two ballooned regions and corresponding rupture openings were observed, with rupture widths of approximately 0.64 mm for both ruptures and rupture lengths of 4.8 mm and 5.6 mm, respectively.

Capps, Nathan [ORNL]↗
DOE OSTI2023

A Plan to Qualify New Fuel for the High Flux Isotope Reactor for Material Minimization

The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) is one of five high power research reactors the Office of Material Management and Minimization (M3) Program, Office of Conversion is working to convert from using highly enriched uranium (HEU) fuel to using low-enriched uranium (LEU) fuel. This effort stems from the primary objective within the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA) to achieve permanent threat reduction by minimizing, and when possible, eliminating weapon-usable nuclear material around the world. Under M3’s Office of Conversion, the U.S. High Performance Research Reactor (USHPRR) Project is pursuing fuel qualification and licensing of LEU fuels to support the high-performance reactors. All high-performance reactors except HFIR will be converted to LEU monolithic uranium-molybdenum alloy fuel. HFIR will be evaluated for conversion to LEU using a uranium silicide fuel, namely, U3Si2-Al dispersion fuel. The mission of the USHPRR Project is to develop the technology needed to reduce, and eventually eliminate, worldwide use of HEU in civilian applications. The goal is to develop the technical means needed to use low enriched uranium (LEU) instead of HEU fuel in research and test reactors without significant penalties in performance, economics, or safety of the reactors. The USHPRR Project has four major elements, called Pillars: Fuel Qualification (FQ) managed at Idaho National Laboratory (INL), Fuel Fabrication (FF) managed at Pacific Northwest National Laboratory (PNNL), Reactor Conversion (RC) managed at Argonne National Laboratory (Argonne), and Cross-Cutting (CC) managed at Savannah River National Laboratory (SRNL). FQ is responsible for the qualification of the fuel type. RC is responsible for supporting reactor conversion analysis and overseeing licensing submittals leading to conversions of domestic reactors to LEU fuel. For the FQ effort, FQ (INL) worked in collaboration with RC (Argonne) and ORNL to develop the plan for the uranium silicide fuel qualification for HFIR. The resulting HFIR Fuel Qualification Plan provides the general approach for the USHPRR team to move the selected uranium silicide fuel design for HFIR conversion through qualification. Authorization and use in HFIR will be approved through the DOE’s Office of Science. Uranium silicide fuel was previously qualified in NUREG-1313 at an approximate maximum heat flux of 1.4 MW/m2 and a maximum fuel section temperature of about 130°C. In addition to the different regulator process utilized by DOE, these upper limits will be exceeded in HFIR; therefore, further testing will be necessary to ensure the fuel can meet HFIR qualification requirements. The HFIR fuel loading may exceed 4.8 gU/cm3 which was determined in the NUREG-1313 safety evaluation to be acceptable for use in non-power NRC-licensed reactors provided there exist no other safety considerations. In addition, the uranium silicide fuel will need to be qualified in a HFIR-specific design. This plan includes the currently available information from the USHPRR Project Functions and Requirements document and expands these requirements to ensure that planned tests have traceable results providing evidence that the requirements have been met. Data collection methods are discussed as well as the process to show that the requirements have been met. This document is designed to provide a pathway for researchers to obtain data necessary and at the appropriate quality level for HFIR fuel qualification

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Thermal conductivity measurement of U-Mo and U-Mo/Al interaction layers generated from in-pile irradiation using the suspended-bridge method

Here, this study presents the first measurement of the individual thermal conductivity of U-7wt.%Mo fuel particles and U-Mo/Al interaction layers (ILs) from in-pile irradiated dispersion fuel plates, using the suspended-bridge method. Nanorods of U-7wt.%Mo fuel and U-Mo/Al ILs were extracted by focused ion beam (FIB), and their microstructures were characterized with transmission electron microscopy (TEM). TEM revealed finely distributed nanobubbles in the U-7wt.%Mo matrix, along with an amorphous structure in the ILs. The thermal conductivity of in-pile irradiated U-7wt.%Mo was approximately 30% lower than that of the unirradiated material, ranging from 6.7 W/m·K at 300 K to 8.5 W/m·K at 380 K. The ILs exhibited even lower thermal conductivity, from 2.1 W/m·K at 300 K to 2.7 W/m·K at 380 K. These reductions, attributed to nanobubbles, fission products, and irradiation-induced point defects, were analyzed through a combination of microstructural characterization and literature-based transport models, which successfully reproduced the observed degradation trends.

42 - ENGINEERING↗
DOE OSTI2022

Non-destructive analysis of swelling in the EMPIrE fuel test

The European Mini-Plate Irradiation Experiment (EMPIrE) was designed to support the development and testing of a coated uranium-molybdenum (U-Mo) dispersion fuel for the conversion of select high-performance research reactors (HPRRs) to utilize low-enriched uranium (LEU). To aid in the development of the coated fuel form, the EMPIrE test included several plate designs and irradiated them in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) at a high meat power density (~21 kW/cm 3 ) and to high fuel particle fission densities (~6.4 × 10 21 fissions/cm 3 ). These conditions mimic the bounding conditions of the BR-2 reactor in Belgium, where a concurrent irradiation experiment was performed, and exceed those previously explored in dispersion U-Mo fuel plates. A local fuel swelling analysis, as determined through high-fidelity, post-irradiation mini-plate profilometry, was used along with statistical methods to non-destructively evaluate the overall performance and separate the effects of convoluted fabrication variables. While some effects observed with this non-destructive analysis were subtle, others had more significant, and possibly competing, effects on the fuel swelling behavior. In closing, these observations will be examined further with destructive examinations to more fully assess them as the fuel design is developed and qualified.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Scoping Study for Fast Flux Testing in the Advanced Test Reactor

The value of fast spectrum reactors remains prominent in the nuclear technology portfolio. The performance of these reactors can be maximized with advancements in nuclear fuel technologies, but development of these technologies is currently held back by lack of fast spectrum test reactors available to the United States. Spectral modification of experiment positions in the thermal spectrum Advanced Test Reactor (ATR) has long been used to support fast reactor fuel development, but these methods have not been progressed to their full potential. This study investigated the use of concentric rings of aluminum-clad fuel plates in ATR flux traps and thermal neutron absorbing filters to increase fast neutron flux on test specimens. This concept was termed the Boosted Energy Advanced Spectrum Test (BEAST). This approach will enable irradiation of advanced fuel designs in prototypic-length fuel pins and representative flux environment to support post irradiation exams, enable transient testing, and produce the type of data that will permit lead test assembly irradiations in true Sodium Fast Reactors (SFRs) when they become available. Neutronic predictions were performed to investigate BEAST design options and thermal hydraulic models were produced to ensure feasibility of BEAST. Two versions were considered based on the geometric limitations of ATR’s small and large flux traps. The small version was found to be preferable due to slightly higher fast flux and fast-to-thermal neutron ratio. Perhaps more influentially, the small flux trap option was also preferred to avoid conflict with ongoing very high temperature reactor fuel irradiation programs in ATR’s large northeast flux trap. The small flux trap option provided less than half the test volume of the large version, but still had adequate volume for seven SFR pins in cross section which could be stacked two-high in ATR’s 1.2m long core to accommodate up to 14 EBR-II size pins. The preference for the small flux trap configuration should be revisited if additional collaborative test programs emerge with the need to irradiate a significant volume of additional specimens. Calculations were performed regarding a lithium deuteride ring to convert thermal neutrons into 14 MeV fusion neutrons. At the time this report was written these calculations were partially complete and it remains to be seen whether the concept would be worth including in BEAST. Given the preference for the small flux trap option, which does not afford enough room for the 14 MeV ring, it was concluded to defer future work on the lithium deuteride ring. This decision could be revisited if fusion material research programs emerge for collaborative testing in BEAST. A cadmium-lined specimen holder design was found to be adequate in filtering thermal neutrons and preferred over other neutron absorbers based on past experience with cadmium baskets. It was acknowledged that cadmium-bearing hardware would become depleted and need to be replaced occasionally, which appeared feasible from a mechanical design perspective. Neutronic studies investigated different enrichment levels in the booster fuel using uranium-molybdenum alloy dispersion fuel which has performed well in past ATR irradiations. Both options were able to drive fuel pins to SFR-like fission heating rates. The high enriched booster fuel option outperformed the low enriched option by ~20% on key metrics including fast flux and fast-to-thermal ratio, but the low enriched option was favored in order to broaden options for potential fuel suppliers. The preferred BEAST design options including cadmium filter with low enriched booster fuel in the small flux trap configuration was predicted to achieve 6.2E14 n/cm2sec fast flux (>0.1 MeV) with a fast-to-thermal ratio of 44.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗