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At least 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Core Physics Characteristics of Extended Enrichment and High Burnup Boiling Water Reactor Fuel

This paper presents the highlights of boiling water reactor (BWR) core physics studies performed at Oak Ridge National Laboratory as part of a series of studies conducted to compare low-enriched uranium (LEU) with LEU+ fuel. The studies analyzed isotopic fuel content, lattice parameters (Phase 1), and core physics (Phase 2) to identify challenges in operation, storage, and transportation for BWRs and pressurized water reactors (PWRs). Because of a lack of publicly available lattice and core designs for modern BWR fuel assemblies and reactor cores, several optimized lattice designs were generated, and different core loading strategies were investigated. Twelve optimized lattice designs with 235 U enrichments ranging from 1.6% to 9% and gadolinia loadings ranging from 3 to 8 wt% were used to model axial enrichment and geometry variations in fuel assemblies for core designs. Each core shares a common set of approximations in design and analysis to allow for consistent comparisons between LEU and LEU+ fuel. The objective is to highlight anticipated changes in core behavior with respect to the reference LEU core. The results of this study show that the differences in LEU and LEU+ core reactor physics characteristics are less significant than the differences in lattice physics characteristics reported in the Phase 1 studies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Fuel Bonding and its Impact on Axial Gas Communication Behavior in Light-Water Reactor Fuel Rods

Axial gas communication concerns the flow along the axial axis of nuclear fuel rods during ramp and loss of coolant accident (LOCA) conditions. During power ramps, the higher linear heat generation rate may cause fuel-to-clad gap closure that may prevent transport of released fission gases to the plenum. Upon reduction in power the gas then can communicate to the plenum. This phenomenon has been experimentally observed by short power dips during ramp experiments completed at the Risø reactor. At higher burnups it is observed that the UO2 fuel and Zircaloy cladding forms a chemical bond. This bond results in complete closure of the gap. When these high burnup rods are subjected to a LOCA, the bond has implications on both the mechanical response (i.e., ballooning) of the cladding and subsequent fuel relocation and axial gas communication. In the LOCA scenario, gas communication is of interest in two different regimes: 1) pre-rupture communication from the plenum towards the lower pressure ballooning area and 2) the post-rupture depressurization of the plenum to the external system pressure. In both regimes the presence of a fuel-to-cladding bond will impact the rate of depressurization. In this work we present a fuel-to-clad bonding model that is coupled to an existing axial gas communication model framework in the BISON fuel performance code. The effect of considering the bond on fuel performance modeling predictions is presented through comparisons to existing experimental data. Experiments considered include several rods from the Halden IFA-650 test series. An evaluation on a full-length rod that explores the combined effect of plenum size and bonding status on axial gas communication behavior is also presented.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Resistance pressure weld for nuclear reactor fuel rod tube end plug

A fuel rod for a nuclear reactor, including a cladding tube having a first end with an annular end face, a second end with an annular end face, and a cylindrical body portion extending therebetween, and a first tube end plug including a front portion, an annular lip with an annular end face, and a substantially straight cylindrical body portion extending therebetween, wherein the surface area of the annular end face of the first end of the cladding tube and the annular end face of the annular lip of the first tube end plug are substantially equal, and the annular end face of the first end of the cladding tube and the annular end face of the annular lip of the first tube end plug are connected by a resistance pressure weld.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Chloride-Based Volatility for Waste Reduction and/or Reuse of Metallic-, Oxide- and Salt-Based Reactor Fuels

The objective of the chloride based volatility project (CBV) was to demonstrate the ability to separate uranium from used fuel to enable process improvements resulting in 10x reduction of waste volume, while maintaining safeguards standards and global backend costs at $\$$1/MW-hr. Current industrial practices perform separations of used fuel using solvent media in the form of aqueous and molten salt processes, resulting in contaminated process waste. The CBV approach utilized solid state chemistry with no solvent media and was successful in chlorinating uranium and fission product oxides that sublimed into the gas phase and were collected in targeted condensation zones based on temperature gradients of chlorinated products. Recovery of better than 95% of initial uranium in the form of UCl4 was demonstrated when simulated used nuclear fuel was used. Laser induced breakdown spectroscopy, LIBS, and ultraviolet-visible spectroscopy, UV-Vis, were combined into a high temperature flow cell design and utilized as process monitoring techniques to observed chlorinated products leave the reaction vessel in real time.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Reevaluating the Thermal-Spectrum Molten Salt Breeder Reactor Fuel Cycle in a Modern Context [Slides]

Thermal-Spectrum Molten-Salt Breeder Reactors (TS-MSBRs) can have highly advantageous characteristics. There is potential for excellent safety, actinide resource utilization, proliferation resistance, waste generation, and exergy characteristics. Realizing advantageous characteristics derives from a combination of proper design/engineering and inherent properties. Grossly unacceptable properties are possible with poor design choices. TS-MSBR designs and technologies remain immature. The United States has not had a focused TS-MSBR program in nearly 50 years. TS-MSBRs have progressed due to: (1) technology development in related areas (fusion, materials, instrumentation); (2) general, worldwide molten-salt reactor (MSR) development activities; (3) molten-salt property measurement and evaluation; and (4) safety evaluation methods and performance-based, technology neutral licensing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Purification of U from U-10Mo scrap generated during the fabrication of high performance research reactor fuel

A low enriched U-Mo alloy fuel is under development to replace highly enriched U fuels currently used in United States high performance research reactors. The alloy casting and fuel fabrication processes will generate scrap streams containing low enriched U (LEU) which must be recovered. Solvent extraction processes were designed using the Argonne Model for Universal Solvent Extraction (AMUSE) to purify solutions containing 20 and 50 g/L U. The feed for the solvent extraction processes was prepared from solutions generated from the dissolution of U-10Mo-Zr foils and U-10Mo-Zr-Al mini-plates. The U purification processes were demonstrated using two, 16-stage banks of miniature mixer-settlers. The solvent extraction experiments demonstrated that all design objectives for the U purification processes could be met. The U recovery in the product stream for each flowsheet was ≥99.9%. The flowsheet demonstrations also showed that the purity of the U Product will meet the requirements of the ASTM International C1462-21 specification for LEU metal enriched to less than 20% 235 U. In conclusion, the AMUSE modeling for both flowsheet demonstrations was validated by comparing predicted and measured concentrations of U, Mo, and Zr in the exit streams and stage samples at steady-state conditions in the mixer-settlers.

modified PUREX process↗
DOE OSTI2024

SCALE Analyses of Scenarios in the High-Temperature Gas-Cooled Reactor Fuel Cycle

This report demonstrates the SCALE code system’s capabilities in modeling and simulating scenarios in the High-Temperature Gas-cooled Reactor (HTGR) nuclear fuel cycle as part of a United States Nuclear Regulatory Commission (NRC) project that aims to demonstrate the capabilities of the SCALE and MELCOR codes for non-light water reactor (non-LWR) modeling and simulation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Transport Modeling of As-Run ATR Cycles to Support U-10Mo Research Reactor Fuel Qualification Experiment

The Department of Energy’s (DOE) Office of Materials Management and Minimization has been tasked with converting the five remaining United States High Performance Research Reactors (U.S. HPRR) from highly enriched uranium to low-enriched uranium. The Nuclear Regulatory Commission (NRC) regulates Massachusetts Institute of Technology Reactor (MITR), Missouri University Research Reactor (MURR), and National Bureau of Standards Reactor (NBSR); and DOE regulates the High Flux Isotope Reactor and Advanced Test Reactor (ATR). To meet the high demands of these reactors, the U.S. HPRR program has chosen to use 90% uranium - 10% molybdenum (U-10Mo) monolithic fuel. This plate-type fuel will undergo multiple irradiation experiment campaigns in ATR, from mini-plates to full element tests, over a large range of operating conditions. This will provide data in support of the fuel qualification of each reactor. This summary focuses on the as-run neutronic analysis of the first series of mini-plate (MP-1) experiments, which have been irradiated in the ATR. MP-1 experiment’s main goal is to demonstrate the fabrication process and meet the fuel irradiation performance requirements with primary focus on the fuel plates associated with the three NRC reactors, with a small focus on the low power requirements of the ATR fuel. The MP-1 experiments are planned to be the basis of the monolithic U-10Mo fuel for qualification through the NRC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Technical Program Plan for INL Advanced Reactor Technologies Advanced Gas Reactor Fuel Development and Qualification Program

High-temperature gas cooled reactors (HTGRs) are graphite moderated nuclear reactors cooled with helium. Their high outlet temperatures and thermal energy conversion efficiency enable efficient and cost effective integration with non electricity generating applications. These applications include process heat and hydrogen production for petrochemical and other industrial processes that require operating temperatures between 300 and 900°C. HTGRs will supplement the use of premium fossil fuels such as oil and natural gas, improve overall energy security in the United States by reducing dependence on foreign fuels, and reduce carbon dioxide (CO2)/greenhouse gas emissions. The HTGR design uses helium as a coolant, graphite as a neutron moderator, and ceramic particle fuel. Helium is chemically inert and neutronically transparent. The graphite core slows down the neutrons, retains its strength at high temperatures, provides structural stability, and acts as a substantial heat sink during transient conditions. The ceramic particle fuel is extremely robust and retains the radioactive by products of the fission reaction within the coated particle under normal and off normal conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Technical Program Plan for INL Advanced Reactor Technologies Advanced Gas Reactor Fuel Development and Qualification Program

High-temperature gas cooled reactors (HTGRs) are graphite moderated nuclear reactors cooled with helium. Their high outlet temperatures and thermal energy conversion efficiency enable efficient and cost effective integration with non electricity generating applications. These applications include process heat and hydrogen production for petrochemical and other industrial processes that require operating temperatures between 300 and 900°C. HTGRs will supplement the use of premium fossil fuels such as oil and natural gas, improve overall energy security in the United States by reducing dependence on foreign fuels, and reduce carbon dioxide (CO2)/greenhouse gas emissions. The HTGR design uses helium as a coolant, graphite as a neutron moderator, and ceramic particle fuel. Helium is chemically inert and neutronically transparent. The graphite core slows down the neutrons, retains its strength at high temperatures, provides structural stability, and acts as a substantial heat sink during transient conditions. The ceramic particle fuel is extremely robust and retains the radioactive by products of the fission reaction within the coated particle under normal and off normal conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Irradiation Performance: High-Temperature Gas Reactor Fuels

The response of TRISO fuel under irradiation is complex and dictated by the integrated nature of all components which define the TRISO fuel architecture. Potential failure mechanisms have been well understood and iterative improvements to the TRISO fuel architecture have been made to ensure optimized performance for different reactor designs. Ultimately, modern TRISO fuel for HTGR applications has exceeded performance targets.

Gerczak, Tyler↗
DOE OSTI2021

Fast Reactor Fuel Testing Considerations

The slide deck is to be presented in a public (i.e., no content restrictions) panel session titled "Versatile Test Reactor: Current Developments" at the 2021 ANS Virtual Annual Meeting. The slides will be used for opening remarks and to present to the audience the technical considerations that motivate the experiment capabilities being incorporated into the Versatile Test Reactor. The line plot included on slide is copied directly from a journal publication, as cited. The values use to make the bar chart come from an open document (ANL-NSE-1) available on OSTI.gov, and are from the EBR-II Mark-VA U-Pu-Zr core design, which was never used. They will be described simply as an unused but illustrative EBR-II core design, and for the purpose of making the point that a range of irradiation and in-service conditions need to be addressed when testing and qualifying a new fuel design for a fast reactor.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Characterization of Fuel and Cladding In and Near the Pellet-Pellet Gap of a High-Burnup Pressurized Water Reactor Fuel Rod

Oak Ridge National Laboratory (ORNL) is performing extensive destructive examinations of 15 high-burnup (HBU) spent nuclear fuel (SNF) rods from the North Anna Power Station (NAPS), which is operated by Dominion Energy Virginia [1]. The examinations are being conducted to obtain a baseline condition of the HBU rods before dry storage and are focused on understanding overall SNF rod strength and durability [2,3]. The HBU rods, referred to as sister rods or sibling rods, are similar to rods placed into dry storage at NAPS that are planned to be examined after one decade. The sister rods include several ZIRLO®-clad rods manufactured by Westinghouse Electric Company. One of the ZIRLO®-clad rods examined includes a pellet-pellet gap of 1 mm that was identified during the nondestructive examinations (NDEs) [4]. The rod was sectioned axially at that elevation to reveal the pellet-pellet interfaces and the pellet-pellet gap. The section was mounted and polished to reveal the distribution of hydride precipitates in the cladding above, below, and within the gap [5]. Total cladding hydrogen measurements will be performed to quantify the total cladding hydrogen through the gap and any additional in-solution or precipitated hydrogen in the pellet-pellet gap region

Montgomery, Rose↗
DOE OSTI2024

Characterization of Fuel and Cladding In and Near the Pellet-Pellet Gap of a High-Burnup Pressurized Water Reactor Fuel Rod

At TopFuel 2021, we reported on metallographic examinations of a high burnup ZIRLO-clad spent fuel rod operated at the North Anna Power Station having a 3-mm gap between pellets at an elevation of ~1,403 mm. The rod was sectioned axially at that elevation to reveal the pellet-pellet interfaces and the pellet-pellet gap. Sections were mounted and polished to reveal the distribution of cladding inner and outer oxide layer thicknesses above, below, and within the gap. Since then, total cladding hydrogen measurements have been performed to quantify the cladding hydrogen content through the gap and above and below the pellet-pellet gap region. This paper provides the cladding hydrogen concentrations and discusses the overall findings regarding cladding waterside oxidation and cladding hydrogen diffusion to a pellet-pellet gap.

Montgomery, Rose↗
DOE OSTI2025

Disruptive thermal-spectrum molten salt breeder reactor fuel cycle technologies

Multiple, disruptive fuel salt processing technologies for thermal-spectrum molten salt breeder reactors (TS-MSBRs) have been conceived of and brought to laboratory levels of maturity since the termination of the historic U.S. government program. This paper describes these technologies and the role that they could serve in decreasing the remaining technical risks for TS-MSBR deployment. In conclusion, the focus of the paper is on describing how the technologies could integrate into a dramatically simpler, proliferation resistant fuel salt processing system as well as identifying remaining development hurdles.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗