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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 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

An integrated approach to examine fuel-cladding chemical interaction in HT9/U-10Zr metallic fast reactor fuels: Coupling machine learning with electron microscopy and local mechanical properties analysis

The metallic U-Zr nuclear fuel alloy has garnered renewed interest as a promising candidate for next-generation sodium-cooled fast reactors. Recent studies and technology assessments have identified several areas requiring improvements, enhanced knowledge, and reliable data to strengthen the U-Zr fuel design basis for qualification and commercial applications. One of the most challenging phenomena impacting this fuel system’s performance is fuel-cladding chemical interaction (FCCI). This work aimed to harvest FCCI data by examining selected HT9/U-10Zr (wt. %) fuel samples of prototypic full-length fuel pins through an integrated approach. This approach integrated scanning electron microscopy (SEM) microstructure characterization with localized mechanical properties examination to deepen understanding of FCCI phenomenon in HT9/U-10Zr fuel system. Particularly, this study focused on MFF fuel pins irradiated at Fast Flux Test Facility (FFTF), which aimed to qualify metallic fuel as a driver fuel for FFTF and to assess its viability for larger-scale fast reactors. Electron microscopy provided high confidence in detecting and distinguishing the different FCCI layers, while small-scale mechanical testing (SSMT) probed the mechanical properties of these layers. SEM examination of a MFF-2 pin 192167, with a time averaged inner cladding temperature (TICT) slightly over 500°C, revealed minimal cladding-side FCCI (cladding wastage). In contrast, significantly thicker cladding wastage comprising two distinct sublayers was observed in samples from the thermally hot MFF-3 pin 193045 and MFF-5 pin 195011 where the TICT ranged from 610-635°C. SSMT indicated complete embrittlement in the sublayer adjacent to the fuel and a tendency toward embrittlement in the other sublayer. Additionally, a new machine learning method was developed, validated, and used to quantify cladding wastage thickness. The machine learning method reliably predicted the wastage thickness across various fuel pins and sample cross-sections. Furthermore, the available cladding wastage data from HT9/U-10Zr fuel system demonstrated a strong temperature dependency. However, the dataset remains small, and ongoing research activities are essential to further understand the FCCI phenomenon and develop a reliable FCCI model for enhanced fuel performance simulation under various conditions.

36 - MATERIALS SCIENCE↗
NASA NTRS1972

Some observations on uranium carbide alloy/tungsten compatibility

Chemical compatibility between both pure and thoriated tungsten and uranium carbide alloys was studied at 1800 C for up to 3300 hours. Alloying with zirconium carbide appeared to widen the homogeneity range of uranium carbide, making additional carbon available for reaction with the tungsten. Reaction layers were formed both by vapor phase reaction and by physical contact, producing either or both UWC2 and W2C, dependent upon the phases present in the starting fuel alloy. Formation of UWC2 results in slow growth of the reaction layer with time, while W2C reaction layers grow rapidly, allowing equilibrium to be reached in less than 2500 hours at 1800 C. The presence of a thermal gradient had no effect on the reactions observed nor did the presence of thoria in the tungsten clad.

Phillips, W. M.↗
NASA NTRS1972

Some observations on uranium carbide alloy/tungsten compatibility.

Results of chemical compatibility tests between both pure tungsten and thoriated tungsten run at 1800 C for up to 3300 hours with uranium carbide alloys. Alloying with zirconium carbide appeared to widen the homogeneity range of uranium carbide, making additional carbon available for reaction with the tungsten. Reaction layers were formed both by vapor phase reaction and by physical contact, producing either or both UWC2 and W2C, depending upon the phases present in the starting fuel alloy. Formation of UWC2 results in slow growth of the reaction layer with time, while W2C reaction layers grow rapidly, allowing equilibrium to be reached in less than 2500 hours at 1800 C. Neither the presence of a thermal gradient nor the presence of thoria in the tungsten clad affect the reactions observed.

Phillips, W. M.↗
DOE OSTI2020

Microstructure and diffusion behavior of uranium fuel with minor additives

Fission product lanthanides in metallic fuels are known to cause adverse fuel-cladding chemical interaction (FCCI). Tin (Sn) and palladium (Pd) are being explored as the potential additives to reduce or mitigate lanthanide-induced FCCI by forming stable Sn-Ln and Pd-Ln compounds. The current study is an investigation of the fuel alloys, U–4Sn, U–4Sn-4Ln, U–4Pd, and U–4Pd-4Ln (wt. %), and their diffusion behaviors with Fe. Microstructural analysis was performed using scanning electron microscopy (SEM). Further, the binary phases in the UPd and USn alloys are identified as UPd 3 and suspected U 2 Sn, where U 2 Sn is unknown in the literature. The binary phases in the UPdLn alloy are identified as PdLn and Pd-rich Pd-Ln (Pd 4 Ln 3 and Pd 3 Ln 2 ), and in the USnLn alloy is SnLn. Diffusion of Fe with these binary phases is insignificant compared to the diffusion with U. The Ln-induced FCCI can be mitigated by the additives.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Formation and Consolidation of UN Nanostructures in U-Mo Fuel via Mechanical Alloying and Spark Plasma Sintering

Uranium-molybdenum (U-Mo) alloys show promise as a nuclear fuel system due to their high thermal conductivity and high fuel loading capability. However, U-Mo systems suffer from irradiation induced swelling ultimately affecting the cladding via mechanical and chemical interaction. To address these shortcomings, this thesis investigated the formation of uranium mononitride (UN) nanoparticles within a U-10Mo matrix. To promote the formation of UN, U-10Mo powders were mechanically alloyed under a high purity nitrogen cover gas (99.9995%) atmosphere. Variations were made in stainless steel milling media size, duration of milling, and number of times the milling jar was re-aerated with nitrogen gas. UN nanoparticulates were successfully formed within the U-Mo matrix and was characterized utilizing light element analysis, X-ray diffractometry, scanning and transmission-electron microscopy, electron energy loss spectroscopy, and atom probe tomography. Presence of the UN nanoparticles was found as early as 1-hour into the mechanical alloying process, and significant iron contamination found after 10-hours of milling. The U-Mo/UN powders were compacted using spark plasma sintering techniques in order to stabilize the nanostructure. U-Mo/UN compacts containing small amounts of neodymium and helium were analyzed to observe the defect migration tendencies of fission products within the fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Pre-Transient Characterization of Historic EBR-II Pins for Transient Testing

Current interest in sodium-cooled fast reactor (SFR) designs, such as TerraPower’s Natrium Reactor, has highlighted the need for advanced reactor fuel technology development. Modern U-Zr and U- Pu-Zr pin designs are primary candidates to fuel SFRs and boast high fuel utilization capacity, increased fuel-cladding compatibility, and improved safety through inherent feedback mechanisms. Despite over 60 years of metallic fuel irradiation, uncertainties exist in the performance of the fuel system, particularly under transient overpower (TOP) and loss of flow (LOF) scenarios. Throughout historical testing within the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF), fuel behavior has demonstrated benign response to transient reactor conditions; however, accurate predictions of failure thresholds to inform operational limitations rely heavily on fuel composition, burnup, and irradiation history. In expanding TOP and LOF testing, the Transient Heat sink Overpower Response (THOR) Capsule will be used to test modern fuel technologies in a static sodium environment in the Transient Reactor Test (TREAT) Facility. The THOR capsule is highly instrumented and will provide time-dependent thermal behavior of SFR fuel pins subjected to accident conditions within TREAT. The THOR-Metallic (THOR- M) campaign aims to validate and expand historical TOP and LOF testing on high burnup U-Zr and U-Pu- Zr fuel alloys previously irradiated in EBR-II by running the rods to failure. This contribution focuses primarily on the pre-transient engineering-scale destructive and non- destructive characterization that has been conducted on both the test and sibling pins used for the TOP and LOF tests. All pins underwent visual examination, neutron radiography, element contact profilometry, and precise gamma scan. The sibling pins used for each test were further analyzed using gas assay, sampling, and recharge analysis (GASR), and optical microscopy. The results from each technique confirmed that the fuel pins were intact and devoid of any atypical developments when compared to historical data. Additionally, the analyzed measurements establish a baseline for comparison to post-transient analysis. Key fuel behaviors quanitifed include axial elongation of the fuel column, diametral strain of the pin, patterns in fluff structure geometry, changes in axial isotope distribution, evolution of constituent redistribution, porosity, and fission gas release. The pre-transient measurements and changes attributed to transient behavior from post-transient measurement will be compared to historical data to capture the behavioral dependence on composition, burnup, and irradiation history. Results from this work advance the initiatives of the THOR-M campaign, which aid in informing fuel performance models and establishing safety criteria for SFR operational limits. The novel combination of test environment, in-situ instrumentation, and comprehensive suite of characterization methods provides greater understanding of transient fuel behavior. Overall, information on the time and condition of pin failure for high burnup U-Pu-Zr will greatly expand the limited existing TOP and LOF test data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Pre-Transient Characterization of Historic EBR-II Pins for Transient Testing

Current interest in sodium-cooled fast reactor (SFR) designs, such as TerraPower’s Natrium Reactor, has highlighted the need for advanced reactor fuel technology development. Modern U-Zr and U- Pu-Zr pin designs are primary candidates to fuel SFRs and boast high fuel utilization capacity, increased fuel-cladding compatibility, and improved safety through inherent feedback mechanisms. Despite over 60 years of metallic fuel irradiation, uncertainties exist in the performance of the fuel system, particularly under transient overpower (TOP) and loss of flow (LOF) scenarios. Throughout historical testing within the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF), fuel behavior has demonstrated benign response to transient reactor conditions; however, accurate predictions of failure thresholds to inform operational limitations rely heavily on fuel composition, burnup, and irradiation history. In expanding TOP and LOF testing, the Transient Heat sink Overpower Response (THOR) Capsule will be used to test modern fuel technologies in a static sodium environment in the Transient Reactor Test (TREAT) Facility. The THOR capsule is highly instrumented and will provide time-dependent thermal behavior of SFR fuel pins subjected to accident conditions within TREAT. The THOR-Metallic (THOR- M) campaign aims to validate and expand historical TOP and LOF testing on high burnup U-Zr and U-Pu- Zr fuel alloys previously irradiated in EBR-II by running the rods to failure. This contribution focuses primarily on the pre-transient engineering-scale destructive and non- destructive characterization that has been conducted on both the test and sibling pins used for the TOP and LOF tests. All pins underwent visual examination, neutron radiography, element contact profilometry, and precise gamma scan. The sibling pins used for each test were further analyzed using gas assay, sampling, and recharge analysis (GASR), and optical microscopy. The results from each technique confirmed that the fuel pins were intact and devoid of any atypical developments when compared to historical data. Additionally, the analyzed measurements establish a baseline for comparison to post-transient analysis. Key fuel behaviors quanitifed include axial elongation of the fuel column, diametral strain of the pin, patterns in fluff structure geometry, changes in axial isotope distribution, evolution of constituent redistribution, porosity, and fission gas release. The pre-transient measurements and changes attributed to transient behavior from post-transient measurement will be compared to historical data to capture the behavioral dependence on composition, burnup, and irradiation history. Results from this work advance the initiatives of the THOR-M campaign, which aid in informing fuel performance models and establishing safety criteria for SFR operational limits. The novel combination of test environment, in-situ instrumentation, and comprehensive suite of characterization methods provides greater understanding of transient fuel behavior. Overall, information on the time and condition of pin failure for high burnup U-Pu-Zr will greatly expand the limited existing TOP and LOF test data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Development and preliminary validation of a mechanistic multiscale model for fuel-cladding chemical interaction in metallic nuclear fuels

Despite decades of fuel rod material and design improvements, fuel-cladding chemical interaction (FCCI) remains the single-most lifetime-limiting behavior for modern metallic fuel rods. Constraining fuel lifetime increases operating costs, limiting the economic viability of commercializing metallic nuclear fuel technology. A mechanistic multiscale model utilizing the finite element method-based MARMOT and BISON codes was developed to more confidently predict cladding-side FCCI and its impact on fuel performance. The new BISON model incorporates mesoscale models for the effects of fuel microstructure evolution on the transport of wastage-inducing lanthanides through the fuel and for the kinetics of cladding wastage layer growth. The mesoscale models, in turn, build on lanthanide transport property data obtained from the atomistic scale. Preliminary validation studies using wastage thickness and cladding profilometry data from four fuel rods irradiated in Experimental Breeder Reactor II experiment X447 and one fuel rod from Fast Flux Test Facility experiment IFR1 show that the new model predicts cladding wastage and its effects on cladding deformation as well as existing empirical FCCI correlations. The new model is expected to aid in the design of new metallic fuel concepts, including fuel additives, cladding liners, and sodium-free annular fuel geometries. In conclusion, future work will focus on broader validation and refinement of the model’s treatment of different fuel alloys and cladding materials.

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 OSTI2023

Godiva IV Burst Reproducibility and Diagnostic Testing

Godiva IV is a fast burst critical assembly located at the Nation Criticality Experiment Research Center (NCERC) in the Nevada National Security Site (NNSS). It is constructed of approximately 65 kg of highly enriched uranium (HEU) fuel alloyed with 1.5% molybdenum for strength. The assembly can be operated at delayed critical or can be used to perform super-prompt critical bursts with temperature rises of up to 250 °C. Several projects with the objective of characterizing Go diva IV are ongoing including characterizations of its radiation emission and thermomechanics. One of the characterization projects is a study of the variability in the relative source emission (or fluence in absolute terms) between bursts that are nominally the same size. There is some inherent variation in the number of fissions, or "size" of the burst, even for bursts where the operators perform identical processes. The source of the variation can be attributed to several factors. The largest contributor is the precision with which the operators are able to determine delayed critical (DC). Operators find DC for every burst operation. The impact of being slightly above or below DC can change the super-prompt critical reactivity and therefore the source term. The temperature of the fuel is another factor. Establishing delayed critical prior to performing a burst act as a control of the bulk temperature coefficient of reactivity.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Swelling and Fission Gas Release of U-10Mo and U-17Mo Following Neutron Irradiation at 250 – 500°C

The performance of metallic fuel alloys, U-10Mo and U-17Mo, was examined using the MiniFuel test system in the High Flux Isotope Reactor. Approximately 0.8 mm thick disks were irradiated at target temperatures of 250°C, 350°C, 450°C, and 500°C up to three different fission densities, culminating at a maximum fission density of 6.8 × 10 20 cm −3 . Fission rates decayed from 3 – 6 × 10 13 cm −3 s −1 to 2 – 4 × 10 13 cm −3 s −1 over the course of the longest, eight cycle, irradiation as the 235 U was consumed and the 239 Pu concentration accumulated. Actual irradiation temperature on the last day of irradiation was measured via dilatometry using the SiC passive thermometry recovered from the MiniFuel subcapsules and compared favorably with the thermal calculations using as-built geometry and test conditions. Furthermore, average simulated temperatures were within 50°C of the target temperatures, except for the 500°C irradiation, for which the temperature variation was 62°C, 32°C, and 90°C in the in two, four, and eight cycle irradiations, respectively. Fission gas release (FGR) measurements showed no release above recoil for any U-17Mo fuels or for the U-10Mo fuel irradiated at 250°C. Finally, significant (40%–80%) FGR was found for the medium- to highest-burnup U-10Mo samples irradiated at target temperatures 350°C–500°C. Significant FGR correlated with sample thickness swelling, which was as high as 13%–35% for high-release samples and below 7% for all other (low–gas release) samples.

FGR↗
DOE OSTI2020

Cold spray technology in nuclear energy applications: A review of recent advances

Cold spray technology is a promising solid-state, powder-based deposition methods for fabrication of coatings, near-net-shape manufacturing, and component repair. In recent years, this technology has been investigated for many applications in the nuclear energy sector. For example, it has been explored for the deposition of coatings of corrosion and oxidation materials on zirconium-alloys fuel claddings in light water reactors (LWR) for achieving more accident-tolerant fuel (ATF) cladding. Other examples of its use for nuclear energy applications include coatings for enhancements in wear resistance and heat transfer, near-net shape manufacturing of next-generation fuel claddings of oxide dispersion-strengthened (ODS) steels, and for mitigation and repair of potential chloride induced stress corrosion cracking in used fuel storage canister systems. This work reviews published literature on the application of cold spray process for nuclear energy applications, while pointing to gaps that have yet to be addressed.

36 MATERIALS SCIENCE↗
DOE OSTI2024

A neutron fluence map of the Los Alamos National Laboratory Godiva IV critical assembly

A neutron fluence map and a total ionizing dose map of the Los Alamos National Laboratory Godiva IV fast burst critical assembly was generated using passive reactor dosimetry, comprised of sulfur pellets and thermoluminescent dosimeters. Godiva IV is an unmoderated, fast burst, critical assembly constructed of approximately 65 kg of highly enriched uranium fuel alloyed with 1.5 % molybdenum for strength. The mapping was performed during a single 75.6 ºC temperature rise burst operation, with the top and sides of the cylindrical Godiva-IV Top Hat covered in passive dosimetry. Dosimetry was placed in a symmetric pattern around the Top Hat, with higher concentrations near the control rods and burst rod. A specific portion of the lower quadrant of the burst rod was mapped to confirm a testing region where the neutron fluence varied by no more than ± 5%. The results will be used to assess the neutron, gamma, and total ionizing dose environment in three-dimensional space around the assembly for higher fidelity experiment placement, active dosimetry positioning, and radiation field characterization.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Radiation characterization summary for the WSMR fast burst reactor environment at the 6-inch location

The characterization of the neutron, prompt gamma-ray, and delayed gamma-ray radiation fields for the White Sands Missile Range (WSMR) Fast Burst Reactor, also known as molybdenum-alloy Godiva (Molly-G) has been assessed at the 6-inch irradiation location. The neutron energy spectra, uncertainties, and common radiation metrics are presented. Code-dependent recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. The Molly-G core was designed and configured similarly to Godiva II, as an unreflected, unmoderated, cylindrical annulus of uranium-molybdenum-alloy fuel with molybdenum loading of 10%. At the 6-inch position, the axial fluence maximum is about 2.4×10 13 n/cm 2 per MJ of reactor energy; about 0.1% of the neutron fluence is below 1 keV and 96% is above 100 keV. The 1-MeV Damage-Equivalent Silicon (DES) fluence is estimated at 2.2×10 13 n/cm 2 per MJ of reactor energy. The prompt gamma-ray dose is roughly 2.5E+03 rad(Si) per MJ and the delayed gamma-ray dose is about 1.3E+03 rad(Si) per MJ.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

A New Era of Nuclear Criticality Experiments: The First 10 Years of Godiva IV Operations at NCERC

The work presented in this paper focuses on the first 10 years (2011–2020) of Godiva IV operations at the National Criticality Experiments Research Center (NCERC). Godiva IV is a fast burst critical assembly constructed of approximately 65 kg of highly enriched uranium fuel alloyed with 1.5% molybdenum for strength. Godiva is one of the last such critical assemblies in the United States and can be used for studies of super-prompt-critical behavior as well as irradiations and demonstrations. An overview of the startup of Godiva IV at NCERC is given followed by a summary of the experiments performed utilizing Godiva IV over the first 10 years of operation at NCERC.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Godiva Experiments for the Nuclear Criticality Safety Program (NCSP)

Godiva IV is a fast burst critical assembly constructed of approximately 65 kg of highly enriched uranium (HEU) fuel alloyed with 1.5 percent molybdenum for strength. Godiva is one of the last such critical assemblies in the United States, and can be used for studies of super-prompt critical behavior as well as irradiations and demonstrations. A demonstration of a Godiva burst is usually performed as a highlight of the hands-on portion of the Criticality Safety Training Classes taught at the National Criticality Experiment Research Center (NCERC). The Godiva burst is used to demonstrate the concept of super-prompt critical and the time-scale of a criticality accident. In addition, several NCSP projects have been conducted on Godiva IV over the past two years. One experiment focused on collecting data to support multiphysics simulations using Photo-Doppler Velocimetry (PDV) to measure surface movement and gamma detectors to measure the burst output as the burst develops from background to peak over ten orders of magnitude. Another experiment was performed to demonstrate the functionality of the Criticality Accident Alarm System (CAAS) system developed for installation in the Y-12 Uranium Processing Facility (UPF). The system must not only respond to a criticality event and alarm, but must also be shown to operate in a high dose environment.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

IER 557: Godiva Pulse Repeatability and Characterization (CED-3A Report)

Godiva IV is a fast burst critical assembly constructed of approximately 65 kg of highly enriched uranium (HEU) fuel alloyed with 1.5% molybdenum for strength. Godiva IV is proposed as the radiation source for IER 498: Design a Shielding Benchmark with a Metal Critical Source. IER 498 proposes a two-burst method to determine the performance of shielding samples, comparing the transmission of the source radiation to detectors with and without a shielding sample present. This eliminates the need for an absolute calibration and allows many uncertainties to cancel. However, the source term variability must be low compared to other sources of experimental uncertainty to ensure the overall uncertainty meets acceptable levels to provide useful measurements. The primary goal of this experiment is to quantify the source term variability, or reproducibility, of Godiva bursts, focusing on the 70 °C bursts planned for IER 498 measurements. Discussions during IER 498 noted that a number of bursts (on the order of six to ten) as identical as possible would be performed and that a 5-7% standard deviation in source intensity would be acceptable.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗