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At least 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Production Facility Simulations for Annular LEU Targets

Eden Radioisotopes, LLC is designing a new molybdenum-99 (Mo-99) production facility based on a TRIGA-sized reactor with a core composed of low enriched uranium (LEU) targets. Reactor production is designed to exceed U.S. demand for Mo-99 using the fission processes. The nominal operational power is 1.8MW (less than 2.0MW, to satisfy the “non-power” reactor requirement) and desired irradiation scenarios include 7, 14, or 21 days of the irradiation for each target element. Safety limits for reactor operating margins are associated with such physical phenomena as fission gas release, fuel swelling, and stress cracking. As shown in the literature, one of the critical parameters defining safety margins for metallic uranium fuel is swelling. In its turn the fuel operational temperature is the main parameter impacting the swelling rate. This report documents the results of consecutive studies of reactor criticality, fission energy depositions, and cooling conditions for the target. The maximum temperature calculated for the LEU targets allows us to choose appropriate modeling approximation of the swelling. All obtained results refer to the preliminary targets and core designs provided by Eden. A criticality analysis was performed, and the results are listed mainly to cross-check Eden’s safety report and assume maximal neutron field disturbance for maximal reactivity. A brief description of the metallic uranium swelling model is provided in the Appendix.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

The University of Missouri Research Reactor (MURR) LEU Fuel Element Flow Test Preliminary Design

As part of the U.S. National Nuclear Security Administration’s (NNSA) mission to eliminate or minimize the civilian use of weapons-grade highly enriched uranium (HEU, ≥ 20 wt% U-235) fuels, the NNSA Material Management and Minimization (M3) Office of Reactor Conversion and Uranium Supply is collaborating with six U.S. high performance research reactors (USHPRR), including one critical facility, to convert from the use of HEU to low-enriched uranium (LEU, < 20 wt% U-235) fuel. Primary conversion objectives for the USHPRR are to develop LEU fuel element designs that will ensure safe reactor operations and maintain the existing experimental facilities performance. The work is being conducted through many interrelated activities by stakeholders across organizations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

HFIR LEU High Density Silicide Dispersion Optimized Design Neutronics Analyses with PHAME

A high-fidelity neutronics model of the Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) with the low-enriched uranium (LEU) high-density silicide dispersion Optimized fuel design was updated and analyzed to generate reactor physics-based metrics to support follow-on thermal hydraulic and transient analyses of this design. The Python HFIR Analysis and Measurement Engine (PHAME) was also updated to enhance the automation capabilities of the framework developed and maintained to perform these reactor physics modeling and simulation efforts. The automated framework significantly increases the efficiency and reproducibility to design and thoroughly analyzes HFIR LEU core designs, changes, and uncertainties. Reactor physics metrics evaluated include but are not limited to fuel depletion, cycle length, fission rate density distributions, axial power peaking factors, kinetics data, reactivity coefficients, control element worths, heat deposition rates, and decay heat. These neutronics results provide essential input to follow-on steady state thermal, thermal hydraulic and reactor transient analyses, which are subject of other reports. The Optimized design operates at 95 MW to maintain HFIR’s current highly enriched uranium core performance level at 85 MW.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Y-12 NATIONAL SECURITY COMPLEX LEU-MO FEEDSTOCK FABRICATION UPDATE

The Y-12 National Security Complex (Y-12) participates in the fuel fabrication pillar of lhe National Nuclear Security Administration's (NNSA's) Office of Material Minimization and Management (M3) Office of Conversion Pillar. Y-12's primary responsibility is to establish a reliable fabrication process of low-enriched uranium-molybdenum (LEU-Mo) feedstock tor United States High Performance Research Reactors (US HPRRs). This update compares US HPRR LEU-Mo feedstock fabrication ettorts over the past several casting campaigns at Y-12 which vary in process methodologies and fabrication parameters. With recent changes to the latest feedstock casting campaign, Process Design Standard-2 (PD-STD-2). Y-12 has seen a significant decrease in isotopic uranium variability of the fuel feedstock fabricated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Reducing DLOFC fuel temperatures by mixing thorium with LEU in a single-zone six-pass fuel cycle in a PBMR-DPP-400 core - 18548

Many studies have been done in neutronics and thermal-hydraulic simulation of the standard 6- pass fuel recirculation scheme for the standard 9.6 wt % enriched, 9 g per fuel sphere low-enriched uranium (LEU) fuel in the PBMR-DPP-400, using different versions of the VSOP diffusion codes. Maximum DLOFC temperatures were all below the upper limit of 1600 deg. C. The DLOFC temperature is highly dependent on the peaking factor of the power; meaning a lower maximum DLOFC temperature can be obtained by suppressing the axial power peak and by moving the radial power peak towards the external reflector. In this study the standard 6-pass fuel recirculation was retained. The improvement strategy was thus attempted by means of flattening the axial power profile by mixing substantial amounts of thorium into the LEU fuel. The addition of thorium led to breading of substantial amounts of {sup 233}U. This led to slower depletion of enrichment with burn-up, which increased fuel reactivity and power densities near the bottom of the core and thus flattened the axial power profile. The effect was a reduction in maximum DLOFC temperature by 44 deg. C. The simulations were made using the VSOP- 99/05 diffusion code. It was further shown that the results obtained are also applicable to the Chinese HTR-PM and the proposed strategies for further improvement can be expected to produce even much better results in Block Reactors than in Pebble Bed Reactors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Fuel element design and analysis for potential LEU conversion of the Advanced Test Reactor

This study provides a summary of the status of the design process for a low-enrichment replacement for the high-enrichment fuel that has historically been used in the Advanced Test Reactor (ATR) at Idaho National Laboratory. Under current long-term Department of Energy policy, the ATR will be converted to operate using low-enrichment fuel by 2030. To this end, an engineering evaluation and design process has been underway for more than ten years. Initially performed as a series of scoping studies for different design options, the last five years have seen a transition to a formal Safety-in-Design approach to the design and safety evaluation process. This paper focuses on such efforts in progress since 2012 to identify pre-conceptual and conceptual design candidates for ATR conversion. The Enhanced LEU Fuel (ELF) design is selected as the most promising conceptual design. Analyses and trade-off studies performed that led to that selection are described. Analysis of certain aspects of core performance for an ELF-fueled core relative to the current high-enrichment fuel show that only small and manageable changes to core operations will be necessary to support conversion of ATR to low-enrichment fuel.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

PWR Core Analysis for Cycle Extension and Uprates with LEU+ Accident Tolerant Fuel and 80 GWd/Tonne Burnup Limit

The U.S. Nuclear Regulatory Commission has recently drafted a rule enabling fuel burnup increase in light water reactors up to 80 GWd/t. In conjunction with use of fuel enrichment up to 10%, and accident tolerant fuel (ATF), this is anticipated to facilitate 24-month cycles in PWRs, along with further power uprates. In this paper, PWR core analysis is performed for 20% increased PWR power output along with cycle extension up to 24 months, in combination with use of chromia-doped fuel and chromium-coated clad, considered to be the most near-term ATF concepts. In combination, these lead to challenging conditions with a core average discharge burnup of up to ~74 GWd/t, challenging even the 80 GWd/t burnup limit. Analysis is performed using the 2-step method with POLARIS (within SCALE) used for the lattice calculations and PARCS for the core calculations. Core designs are first baselined for current operating conditions (LEU, 62 GWd/t discharge burnup limit) and then derived that meet cycle constraints on power distribution and the updated lead pin discharge burnup limit while maintaining at least two batches of fuel in the core. Gadolina loadings in fuel pins of up to 8% are used, with enrichment zoning both within the core and, to a limited extent, within assemblies. Here, doped fuel with coated cladding can utilize the same core designs as the reference UOX cores, exhibiting slightly lower burnup due to higher fuel density, which also offsets the slight reactivity penalty from the doping and coating. For the analysis performed here, doped fuel enabled a core with 24-month cycle and 20% uprate to stay within the 80 GWd/t lead pin discharge burnup limit.

LEU+↗
DOE OSTI2019

Bias Between ENDF/B-VIII and ENDF/B-VII.1 for LEU Pin Array Systems

Extensive testing of new cross section data is performed in SCALE prior to its release. The release of ENDF/B-VIII.0 requires the generation and testing of new data libraries which are planned to be included in the release of SCALE 6.3 in 2019 or 2020. KENO benchmark results have been published previously for beta and final releases of ENDF/B-VIII.0 using the entire Verified, Archived Library of Inputs and Data (VALID). This work presents an energy-dependent change in the results for low-enriched uranium (LEU) pin array systems. The process of identifying the difference is highlighted as a case study to determine how data testing and benchmarking can be performed to provide more detailed feedback to the data community on potential data deficiencies.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

White Paper - Comparison of LEU and HEU Fuel for the Kilopower Reactor

This white paper presents a comparison of Low Enriched Uranium (LEU) versus Highly Enriched Uranium (HEU) as fuel for the Kilopower space reactor. It draws upon a previously published white paper 1 by the LANL authors, which was written in order to explore synergistic programmatic objectives with the current NASA Nuclear Thermal Propulsion (NTP) program. In that white paper no recommendations or conclusions were drawn on the appropriate choice of fuel for Kilopower. However, recent news articles, congressional remarks and conference papers have suggested the use of HEU for Kilopower may need to be reconsidered. This white paper explores in more detail the pros and cons of both fuel choices and presents a conclusion on why the use of HEU is the preferred choice as fuel for Kilopower.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Fabrication and Characterization of Candidate Alloys for Advanced LEU Fuel Concepts

A set of novel uranium-based alloys has been proposed for use in advanced LEU fuel systems. This study seeks to develop methods to fabricate single-phase alloys and evaluate the compatibilities of the alloys for use in composite fuel systems. The as-cast microstructure of each alloy was characterized via scanning electron microscopy (SEM), and candidate alloys were further down-selected for homogenization and interaction testing. The U 22 Zr 20 Nb 58 alloy was successfully fabricated as a single-phase system, as demonstrated via X-ray diffraction (XRD) and SEM. The thermal and mechanical properties of the U-Nb-Zr alloy were further assessed via differential scanning calorimetry (DSC), laser flash analysis (LFA), and Vickers’ micro-hardness testing. This alloy also demonstrated improved compatibility with fuel candidate alloys from conventional matrix materials and may be a viable option for use in advanced fuel systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Characterization of Preliminary Powder Production and Consolidation for Advanced LEU Fuel Concepts

A set of novel uranium-based alloys has been proposed for use in advanced LEU fuel systems that utilize powder metallurgical methods for fabrication of the fuel forms. This study seeks to evaluate the effectiveness of current fabrication methods to produce powder of each of the candidate alloys and consolidate composite systems. Powder samples were produced via atomization and analyzed for particle size, shape, and microstructure utilizing sieving methods and scanning electron microscopy (SEM). Preliminary pellet consolidation tests were conducted with the atomized powder using a press and sinter method. The resulting pellets were analyzed via SEM to evaluate the morphology of the composite system. These preliminary tests demonstrate the ability to produce powder but did not yield the desired quality of powder and impacted the ability to fabricate a fully dense pellet. A refined process is under development to produce a higher yield of small (<250 µm) spherical powder.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Development of Powder Production Methods for Advanced LEU Fuel Concepts

A set of novel fuel concepts has been proposed for use in advanced low-enriched uranium (LEU) systems that utilize powder metallurgical methods for fabrication of the fuel forms. Preliminary tests demonstrated the ability to produce powder but did not yield the desired quality to be used as feedstock in these applications. This study seeks to establish an improved powder production method for uranium-based alloys and evaluate the parameters required to refine the particles produced by size and morphology. Powder samples were fabricated via atomization and analyzed for particle morphology utilizing sieving methods and scanning electron microscopy (SEM). Surrogate testing displayed improved performance in producing small (<250 µm) spherical particles, and initial tests demonstrated the initial capability to atomize uranium. Further work is required to refine the atomization process to produce high quality uranium microspheres.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

US Progress on Property Characterization to Support LEU U-10 Mo Monolithic Fuel Development

The US High Performance Research Reactor program is pursuing development and qualification of a new high density monolithic LEU fuel to facilitate conversion of five higher power research reactors located in the US (ATR, HFIR, NBSR, MIT and MURR). In order to support fabrication development and fuel performance evaluations, new testing capabilities are being developed to evaluate the properties of fuel specimens. Residual stress and fuel-cladding bond strength are two characteristics related to fuel performance that are being investigated. In this overview, new measurement capabilities being developed to assess these characteristics in both fresh and irradiated fuel are described. Progress on fresh fuel testing is summarized and on-going hot-cell implementation efforts to support future PIE campaigns are detailed. It is anticipated that benchmarking of as-fabricated fuel characteristics will be critical to establishing technical bases for specifications that optimize fuel fabrication and ensure acceptable in-reactor fuel performance.

Characterization↗
DOE OSTI2019

LEU-Mo Casting Update

The Y-12 National Security Complex (Y-12) participates in the Fuel Fabrication Pillar of the National Nuclear Security Administration’s (NNSA’s) Office of Material Minimization and Management (M3) Office of Conversion Pillar. Y-12’s primary responsibility is to establish the fabrication process for the low-enriched uranium-molybdenum (LEU-Mo) feedstock.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Status of LEU U-10Mo Monolithic Fuel Testing and Qualification Efforts to Support US High Performance Research Reactor Conversions

The US High Performance Research Reactor Project has begun a new phase of fuel testing and post irradiation examination (PIE) of plate-type low-enriched uranium (LEU) U-10Mo Monolithic fuel. The testing campaigns are intended to first confirm that the fuel performance behavior of commercially produced fuel matches that of the lab-scale fabricated fuel tested during development, followed by qualification of the U-10Mo monolithic fuel system through the US Nuclear Regulatory Commission. Planned tests include miniature-plate, full-size plate, and fuel element irradiations, during which fuel test specimens fabricated with qualified processes will be irradiated under conditions that bound the NRC-regulated reactors in order to demonstrate that the fuel meets the overarching fuel performance requirements of 1) maintaining mechanical integrity 2) maintaining geometric stability and 3) exhibiting stable and predictable behavior. The MP-1 test, with the objective of confirming fuel performance behavior, is currently undergoing irradiation in the Advanced Test Reactor and PIE in INL hot cells. Status of upcoming qualification and licensing test design activities and highlights of initial as-fabricated characterization and testing results from the MP-1 test will be provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

A cellular automaton method to simulate the microstructure and evolution of low-enriched uranium (LEU) U–Mo/Al dispersion type fuel plates

Low-enriched uranium (LEU) fuel plates for high power materials test reactors (MTR) are composed of nominally spherical uranium-molybdenum (U-Mo) particles within an aluminum matrix. Fresh U-Mo particles typically range between 10 and 100 mu m in diameter, with particle volume fractions up to 50%. As the fuel ages, reaction-diffusion processes cause the formation and growth of interaction layers that surround the fuel particles. The growth rate depends upon the temperature and radiation environment. The cellular automaton algorithm described in this paper can synthesize realistic random fuel-particle structures and simulate the growth of the intermetallic interaction layers. Examples in the present paper pack approximately 1000 particles into three-dimensional rectangular fuel structures that are approximately 1 mm on each side. The computational approach is designed to yield synthetic microstructures consistent with images from actual fuel plates and is validated by comparison with empirical data on actual fuel plates. (C) 2014 Elsevier B.V. All rights reserved.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗