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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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6 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

THOR-aLEU Report

The advanced low enriched uranium (aLEU)-THOR experiment vehicle is designed to irradiate fresh fuel rodlet specimens containing UO 2 (uranium dioxide) in the Temperature Heat-sink Overpower Response (THOR) capsule. The goal of the experiments are to compare the behaviors of standard UO 2 pellets to an aLEU concept in which molybdenum (Mo) foils separate UO 2 wafers increasing the thermal conductivity of the fuel. This experiment is designed to be conducted in the Transient Reactor Test facility (TREAT) using the Minimal Activation Retrievable Capsule Holder (MARCH) irradiation system. The goal of these experiments is to assess the effect of thermally conductive inserts in UO 2 by measuring their net effect on radial thermal conductivity using transient nuclear heating and then to determine their power-to-melt threshold in transient overpower ramps. The objective of the aLEU-THOR fuels experiment is to determine the viability and study the performance of UO 2 fuel pellets employing newly upgraded TREAT capability. A novel transient measurement technique will be used with the THOR capsule, which has a solid heat sink to generate the transient temperature gradient for conductivity measurement.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Measured Thermal and Fast Neutron Fluence Rates for aLEU Holders During Cycle 169A

This report contains the thermal (2200 mis) and fast (E> IMeV) neutron fluence rate data for the aLEU holders located in core for ATR Cycle 169A which were measured by the Radiation Measurements Laboratory (RML) as requested by the Power Reactor Programs (ATR Experiments) Radiation Measurements Work Order. This report contains measurements of the fluence rates corresponding to the particular elevations relative to the 80-ft. core elevation. The data in this report consist of (1) a table of the ATR power history and distribution, (2) a hard copy listing of all thermal and fast neutron fluence rates, and (3) plots of both the thermal and fast neutron fluence rates. All "BR" holder monitor wires for this cycle are 56.375 inches long. This length allows measurements from 31.17 inches above core midplane to 19.81 inches below core midplane. The distance from the end of the wires to the first count position was 4.25 inches for all wires counted from this cycle.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Fabrication and Characterization of Single-Particle Compacts

The programmatic objective of the advanced low enriched uranium (aLEU) fuel Project within the Office of Nonproliferation Research and Development (NA-22) is to advance the development of nuclear reactor designs, fuel materials, and fabrication technologies capable of meeting a number of challenging customer requirements, including reactor stability, extended lifetime, and power density without refueling, while using only low-enriched uranium (i.e., less than 20% enrichment of 235 U). The tandem requirements of delivering sufficient power density and maintaining fuel performance over the course of a multidecade fuel lifetime tightly constrain the field of potential fuel options. Monolithic fuel forms (e.g., fuel pellets) provide maximized density of fissile material, but they may face challenges in maintaining their structure and performance over the long reactor lifetimes envisioned in this program. Dispersion fuel forms with an inert matrix provide enhanced long-term stability but sacrifice fissile material density, and they may not provide sufficient power density over the reactor lifetime with the limit of low enrichment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Development of Instrumented Advanced Test Reactor Irradiation Capsule Experiment for In-situ Thermal Conductivity Measurements of High-Density Fuels

Idaho National Laboratory (INL) is developing a first-of-a-kind leadout instrumented capsule experiment design to enable in-situ measurement capabilities in the Advanced Test Reactor (ATR) core. The Ceramic Advanced Thermal Evolution Research (CRATER) experiment supports the aLEU program objective to accelerate fuel performance irradiation testing for identifying alternative high-assay, low enriched uranium (HALEU) fuel systems. CRATER is a fueled, instrumented capsule experiment to measure in-situ temperature and thermal conductivity of ceramic fuels. Two ceramic fuel types will be used, uranium mono-nitride (UN) and uranium mono-carbide (UC), with a third metallic fuel used for comparison (UMo). The three fuel specimens will use a stainless-steel cladding. Programmatic objectives include linear heat generation rates (LHGR) of 210 ± 25 Watts per cm. and an inner clad temperature of 300-450 °C. The evolution of fuel thermal conductivity during irradiation has never been successfully measured in-situ for these systems and this experiment is designed to use advances in measurement sciences to characterize how thermal transport properties evolve while in reactor. Neutronic simulations of the experiment and its surrounding reactor environment were conducted using the Monte Carlo N-Particle Transport code (MCNP) and result in optimized fuel enrichment to meet target linear heat generation rates (LHGRs) influencing fuel temperatures, and fuel burnup requirements. Fabrication research and development (R&D) efforts are underway to produce annular right cylinder UC and UN pellets using carbothermic reduction and nitridation (or hydride-dehydride-nitride) synthesis methods, followed double-action die cold isostatic pressing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Microstructural characterization of as-fabricated monolithic plates with boron carbide, aluminum boride, and zirconium boride burnable absorbers

The use of burnable absorbers can be beneficial for nuclear reactors by extending the fuel’s operational cycle, providing additional criticality control, and flattening the power profile. In this work, three burnable absorber materials (boron carbide, aluminum boride, and zirconium boride) embedded in aluminum have been fabricated into foils and clad in AA-6061 for potential use in high performance research reactors. The as-fabricated boron-containing phases were determined using transmission electron microscopy to be AlB2, B4C, and ZrB2. TEM also revealed weak bonding at the B4C-matrix interface. SEM showed a relatively uniform spatial distribution of boron-containing phases for all the candidate materials. Higher porosity was observed in the foil containing ZrB2 in its as-rolled condition. The porosity in the ZrB2 foil was reduced by hot isostatic pressing. The size and shape distributions of the boron-containing phases were analyzed on the criteria of cross-sectional area, perimeter, roundness, circularity, and aspect ratio. A method of converting the 2D burnable absorber dispersoids seen in cross-sectional microscopy images into 3D volumes was derived using both spherical and ellipsoidal geometry models. The difference in calculated burnable absorber dispersoid average volume between the two models ranges from 20% to 100%, which could impact burnable absorber burnout rates due to differences in neutron self-shielding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Fuel Specification for Uranium Monocarbide FAST Fuel Specimens

This specification outlines the fabrication requirements for kernel compact uranium monocarbide (UC) fuel to be fabricated at General Atomics (GA) and is already issued in EDMS and is going thru LRS so that it can be shared with the Vendor. The UC fuel will be irradiated in the Advanced Test Reactor (ATR) using the Irradiation System for High-Throughput Acquisition (ISHA-1). The primary objective of the ISHA-1 capsule design is to deliver to Idaho National Lab (INL) a semi-universal drop-in capsule that can facilitate the irradiation of both fissile and structural materials in a variety of ATR positions. This irradiation experiment will support the need for development of Accelerated Fuel Qualification (AFQ) methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗