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At least 559 records · Page 31
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Stochastic optimization of a uranium oxide reaction mechanism using plasma flow reactor measurements

Abstract In this work, a coupled Monte Carlo Genetic Algorithm (MCGA) approach is used to optimize a gas phase uranium oxide reaction mechanism based on plasma flow reactor (PFR) measurements. The PFR produces a steady Ar plasma containing U, O, H, and N species with high temperature regions (3000–5000 K) relevant to observing UO formation via optical emission spectroscopy. A global kinetic treatment is used to model the chemical evolution in the PFR and to produce synthetic emission signals for direct comparison with experiments. The parameter space of a uranium oxide reaction mechanism is then explored via Monte Carlo sampling using objective functions to quantify the model-experiment agreement. The Monte Carlo results are subsequently refined using a genetic algorithm to obtain an experimentally corroborated set of reaction pathways and rate coefficients. Out of 12 reaction channels targeted for optimization, four channels are found to be well constrained across all optimization runs while another three channels are constrained in select cases. The optimized channels highlight the importance of the OH radical in oxidizing uranium in the PFR. This study comprises a first step toward producing a comprehensive experimentally validated reaction mechanism for gas phase uranium molecular species formation.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Machine learning surrogates for surface complexation model of uranium sorption to oxides

Abstract The safety assessments of the geological storage of spent nuclear fuel require understanding the underground radionuclide mobility in case of a leakage from multi-barrier canisters. Uranium, the most common radionuclide in non-reprocessed spent nuclear fuels, is immobile in reduced form (U(IV) and highly mobile in an oxidized state (U(VI)). The latter form is considered one of the most dangerous environmental threats in the safety assessments of spent nuclear fuel repositories. The sorption of uranium to mineral surfaces surrounding the repository limits their mobility. We quantify uranium sorption using surface complexation models (SCMs). Unfortunately, numerical SCM solvers often encounter convergence problems due to the complex nature of convoluted equations and correlations between model parameters. This study explored two machine learning surrogates for the 2-pK Triple Layer Model of uranium retention by oxide surfaces if released as U(IV) in the oxidizing conditions: random forest regressor and deep neural networks. Our surrogate models, particularly DNN, accurately reproduce SCM model predictions at a fraction of the computational cost without any convergence issues. The safety assessment of spent fuel repositories, specifically the migration of leaked radioactive waste, will benefit from having ultrafast AI/ML surrogates for the computationally expensive sorption models that can be easily incorporated into larger-scale contaminant migration models. One such model is presented here.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Photocatalytic decomposition of Rhodamine B on uranium-doped mesoporous titanium dioxide

Mesoporous uranium-doped TiO 2 anatase materials were studied to determine the influence of U-doping on the photocatalytic properties for Rhodamine B (RhB) degradation. The physico-chemical properties of the samples were characterized and the results of X-ray diffraction, transmission electron microscopy, and Raman spectroscopy demonstrate homogeneous incorporation of uranium into the anatase lattice. X-ray photoelectron spectroscopy of the doped anatase confirmed the dominance of the U 4+ species and an increasing proportion of U 6+ species as the uranium doping was increased. The absorption thresholds of the uranium-doped anatase extended into the visible light region. A synergistic effect of the band gap energy and oxidation state of the dopant contribute to an enhanced photocatalytic capability for RhB degradation by U-doped TiO 2 .

36 MATERIALS SCIENCE↗
DOE OSTI2020

Isocyanide adducts of tri- and tetravalent uranium metallocenes supported by tetra(isopropyl)cyclopentadienyl ligands

In this work, reaction of the uranium( III ) metallocenium salt [(Cp iPr4 ) 2 U][B(C 6 F 5 ) 4 ] with tert-butyl isocyanide ( t BuNC) yielded the dicationic uranium( IV ) complex [(Cp iPr4 ) 2 U(CN t Bu) 4 ][B(C 6 F 5 ) 4 ] 2 (1), which displays a linear metallocene geometry. Use of crude mixtures of [(Cp iPr4 ) 2 U][B(C 6 F 5 ) 4 ], which contain a soluble source of iodide, led instead to isolation of the monocationic uranium( IV ) iodide complex [(Cp iPr4 ) 2 U(I)(CN t Bu) 2 ][B(C 6 F 5 ) 4 ] (2). Adduct formation with no change in oxidation state was observed upon addition of t BuNC to the neutral uranium( III ) species (Cp iPr4 ) 2 UI, resulting in isolation of (Cp iPr4 ) 2 U(I)(CN t Bu) (3). X-ray crystallographic and IR spectroscopic studies both showed effects ascribed to the presence of multiple strongly donating isocyanide ligands in 1.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Multiple oxidation states of uranium stabilized by an O , N , O -ligand

Elements of the 5f row maintain a wide variety of oxidation states that have been exploited in synthesis, catalysis, and separations. Herein, we describe the complexation of an O,N,O chelator, ExPh, with uranium in the (IV) and (VI) (uranyl) oxidation states. These two uranium complexes, U(IV)ExPh and UO 2 ExPh, respectively, were characterized in the solid state via single-crystal X-ray diffraction (SC-XRD) analysis. Electrochemical studies using cyclic voltammetry were employed to investigate the multiple redox events associated with the U(IV) complex. Spectro-electrochemical analysis of U(IV)ExPh provided spectroscopic evidence of a stable U(V) species. Chemical oxidation of U(IV)ExPh allowed isolation of the U(V) complex, U(V)ExPh. All three stable uranium complexes produced in this study were characterized via IR, UV-vis and NMR spectroscopies, and micro-spectrophotometry. On the other hand, efforts to reduce U(IV)ExPh to the corresponding U(III) species or produce this putative complex directly from ExPh failed to yield an isolable product. The stabilization of three formal oxidation states of uranium, coupled with previous lanthanide-row results, paves the way for studies of ExPh and its analogues in minor actinide chemistry.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

Molecular dynamics simulation of thermal transport in UO 2 containing uranium, oxygen, and fission-product defects

Uranium dioxide (UO 2 ) is the most commonly used fuel in light-water nuclear reactors and thermal conductivity controls the removal of heat produced by fission, thereby governing fuel temperature during normal and accident conditions. The use of fuel performance codes by the industry to predict operational behavior is widespread. A primary source of uncertainty in these codes is thermal conductivity, and optimized fuel utilization may be possible if existing empirical models are replaced with models that incorporate explicit thermal-conductivity-degradation mechanisms during fuel burn up. This approach is able to represent the degradation of thermal conductivity due to each individual defect type, rather than the overall burn-up measure typically used, which is not an accurate representation of the chemical or microstructure state of the fuel that actually governs thermal conductivity and other properties. To generate a mechanistic thermal conductivity model, molecular dynamics (MD) simulations of UO 2 thermal conductivity including representative uranium and oxygen defects and fission products are carried out. These calculations employ a standard Buckingham-type interatomic potential and a potential that combines the many-body embedded-atom-method potential with Morse-Buckingham pair potentials. Potential parameters for UO 2+x and ZrO 2 are developed for the latter potential. Physical insights from the resonant phonon-spin-scattering mechanism due to spins on the magnetic uranium ions are introduced into the treatment of the MD results, with the corresponding relaxation time derived from existing experimental data. High defect scattering is predicted for Xe atoms compared to that of La and Zr ions. Uranium defects reduce the thermal conductivity more than oxygen defects. For each defect and fission product, scattering parameters are derived for application in both a Callaway model and the corresponding high-temperature model typically used in fuel-performance codes. The model is validated by comparison to low-temperature experimental measurements on single-crystal hyperstoichiometric UO 2+x samples and high-temperature literature data. Furthermore, this work will enable more accurate fuel-performance simulations and will extend to new fuel types and operating conditions, all of which improve the fuel economics of nuclear energy and maintain high fuel reliability and safety.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Analysis of Rare Earth Elements in Uranium Using Handheld Laser-Induced Breakdown Spectroscopy (HH LIBS)

In this paper, a portable handheld laser-induced breakdown spectroscopy (HH LIBS) instrument was evaluated as a rapid method to qualitatively analyze rare earth elements in a uranium oxide matrix. This research is motivated by the need for development of a method to perform rapid, at-line chemical analysis in a nuclear facility, particularly to provide a rapid first pass analysis to determine if additional actions or measurements are warranted. This will result in the minimization of handling and transport of radiological and nuclear material and subsequent exposure to their associated hazards. In this work, rare earth elements (Eu, Nd, and Yb) were quantitatively spiked into a uranium oxide powder and analyzed by the HH LIBS instrumentation. This method demonstrates the ability to rapidly identify elemental constituents in sub-percent levels in a uranium matrix. Preliminary limits of detection (LODs) were determined with values on the order of hundredths of a percent. Validity of this methodology was explored by employing a National Institute of Standards and Technology (NIST) standard reference materials (SRM) 610 and 612 (Trace Elements in Glass). Finally, it was determined that the HH LIBS method was able to clearly discern the rare earths elements of interest in the glass or uranium matrices.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Radon Exposure, IL-6 Promoter Variants, and Lung Squamous Cell Carcinoma in Former Uranium Miners

Background: High radon exposure is a risk factor for squamous cell carcinoma, a major lung cancer histology observed in former uranium miners. Radon exposure can cause oxidative stress, leading to pulmonary inflammation. Interleukin-6 (IL-6) is a pro-carcinogenic inflammatory cytokine that plays a pivotal role in lung cancer development. Objectives: We assessed whether single nucleotide polymorphisms (SNPs) in the IL6 promoter are associated with lung cancer in former uranium miners with high occupational exposure to radon gas. Methods: Genetic associations were assessed in a case–control study of former uranium miners (242 cases and 336 controls). A replication study was performed using data from the Gene Environment Association Studies (GENEVA) Genome Wide Association Study (GWAS) of Lung Cancer and Smoking. Functional relevance of the SNPs was characterized using in vitro approaches. Results: We found that rs1800797 was associated with squamous cell carcinoma in miners and with a shorter time between the midpoint of the period of substantial exposure and diagnosis among the cases. Furthermore, rs1800797 was also associated with lung cancer among never smokers in the GENEVA dataset. Functional studies identified that the risk allele was associated with increased basal IL-6 mRNA level and greater promoter activity. Furthermore, fibroblasts with the risk allele showed greater induction of IL-6 secretion by hydrogen peroxide or benzo[a]pyrene diolepoxide treatments. Conclusions: An IL6 promoter variant was associated with lung cancer in uranium miners and never smokers in two external study populations. Lastly, the associations are strongly supported by the functional relevance that the IL6 promoter SNP affects basal expression and carcinogen-induced IL-6 secretion

60 APPLIED LIFE SCIENCES↗
DOE OSTI2014

Analysis of a Uranium Oxide Sample Interdicted in Slovakia (FSC 12-3-1)

We provide a concise summary of analyses of a natural uranium sample seized in Slovakia in November 2007. Results are presented for compound identification, water content, U assay, trace element abundances, trace organic compounds, isotope compositions for U, Pb, Sr and O, and age determination using the 234 U – 230 Th and 235 U – 231 Pa chronometers. The sample is a mixture of two common uranium compounds - schoepite and uraninite. The uranium isotope composition is indistinguishable from natural; 236 U was not detected. The O, Sr and Pb isotope compositions and trace element abundances are unremarkable. The 234 U – 230 Th chronometer gives an age of 15.5 years relative to the date of analysis, indicating the sample was produced in January 1997. A comparison of the data for this sample with data in the Uranium Sourcing database failed to find a match, indicating the sample was not produced at a facility represented in the database.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2018

Historical Review of Stress Corrosion Cracking in Concentrated Uranium-Molybdenum Alloys

This report presents a historical overview of stress corrosion cracking (SCC) in uranium-molybdenum alloys, focusing primarily on U-10 wt. pct. Mo (U-10Mo). This alloy has been chosen as the high density uranium monolithic fuel for converting research reactors from high enriched uranium (HEU) to low enriched uranium (LEU) fuel for its combination of fuel density and gamma phase stability. One disadvantage of U-10Mo is its propensity towards stress corrosion cracking (SCC), driven by high residual stresses resulting from its high yield strength. This report contains a metallurgical overview of U-10Mo alloys incorporating processing, microstructural evolution, and mechanical properties. This report also summarizes SCC literature including contributing factors, initiation and propagation, the mechanism responsible for U-10Mo SCC, and possible solutions to the problem. A brief synopsis of interviews with key personnel from several Department of Energy (DOE) facilities and BWX Technologies (BWXT) is also included. Recommendations for future studies are outlined.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Historical Review of Stress Corrosion Cracking in Concentrated Uranium-Molybdenum Alloys

This report presents a historical overview of stress corrosion cracking (SCC) in uranium-molybdenum alloys, focusing primarily on bare U-10 wt. pct. Mo (U-10Mo). The fuel system being pursued by the United States High Performance Research Reactor (USHPRR) project incorporates a zirconium interlayer applied early in the process which likely serves as a protective coating. This alloy has been chosen as the high density uranium monolithic fuel for converting research reactors from high enriched uranium (HEU) to low enriched uranium (LEU) fuel for its combination of fuel density and gamma phase stability.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Spectrophotometers for Uranium Monitoring in H Canyon

SRNL Analytical Development and R&D Engineering have developed an On-Line Monitoring Spectrophotometer to provide measurements of uranium and nitric acid at several locations in H Canyon Second Uranium Cycle. Instruments are installed at Tanks 15.4, 16.8, 17.4, 17.5, and 18.7. Two-sigma uncertainties for the instruments are 5.3% uranium (all tanks) and 11% nitric acid (Tanks 16.8 and 17.5). The spectrophotometers are of similar design to those recently installed in HB-Line for Pu monitoring, although some changes have been made to make them more suitable for uranium measurements. A full description of design, components, and operating procedures is given.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Environmental Assessment for Use of DOE-Owned High-Assay Low-Enriched Uranium Stored at Idaho National Laboratory

The U.S. Department of Energy (DOE) proposes to make available about 10 Metric Tons (MT) of High-Assay Low-Enriched Uranium (HALEU ) feedstock produced through the electrometallurgical treatment (EMT) process, and other small quantities of HALEU stored at Idaho National Laboratory (INL) for research development & demonstration in support of the commercial nuclear industry and government agencies, including use in advanced reactors. HALEU is a term applied to uranium that is enriched in the uranium-235 (U-235) isotope to a value that is 5% to 20% of the total uranium. Private sector advanced nuclear reactor designs and advanced nuclear fuel designs call for use of HALEU, but currently no commercial facility manufactures HALEU. DOE proposes to expand the fuel fabrication capability at INL to produce HALEU fuel at INL from 10 MT of HALEU feedstock to meet near-term needs. The production requires expansion of the fuel fabrication capability, including the purchase of new equipment and use of facilities at INL’s Materials and Fuels Complex (MFC) and the Idaho Nuclear Technology and Engineering Center (INTEC). DOE has made no decision on the specific use of the fuel. The fuel could be used in advanced reactor design or for the purpose of research and development. DOE would work with other Federal agencies and commercial vendors to determine use of HALEU fuel. Use of this HALEU, including its potential use in a future nuclear reactor, will be analyzed in future National Environmental Policy Act (NEPA) documents by the appropriate agency. The EMT process in operation at the INL converts sodium-bonded spent nuclear fuel into waste forms suitable for disposal as high-level waste and a HALEU product that is unsuitable for diversion to nuclear weapons but could be either reused in fuel or disposed of as waste. The Final Environmental Impact Statement for the Treatment and Management of Sodium-Bonded Spent Nuclear Fuel, DOE/EIS-0306, DOE 2000) discussed the EMT process, but did not make a decision on the disposition or use of the HALEU product from the EMT process. This Environmental Assessment (EA) addresses the HALEU product as HALEU feedstock in production of HALEU fuel.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2019

Stage B Uranium Sequestration Amendment Delivery Monitoring Using Time-Lapse Electrical Resistivity Tomography

The Hanford Site 300 Area lies adjacent to the Columbia River, approximately 5 km north of Richland, Washington. Past waste disposal practices in the 300 Area resulted in vadose zone uranium contamination beneath former infiltration ponds and trenches. Stage-driven water table fluctuations and river water intrusion facilitate mobilization of uranium from contaminated sediments in regions periodically occupied by the water table (i.e., the periodically rewetted zone or PRZ), thereby raising groundwater uranium concentrations above the maximum allowable contaminant level for uranium. This report documents the ERT imaging operations and interpretation of imaging results in terms of the extent of amendment delivery (including phosphate and/or carrier fluid) within the treatment area. Based on the interpretation, amendment delivery to the LVZ and PRZ was observed to be variable between each of the three imaging clusters, as shown in Figure S.1. Amendment transport in Cluster 1 exhibited significant lateral flow in comparison to Clusters 2 and 3, resulting in a final amendment distribution throughout a large portion of the LVZ and PRZ sediment between the injection wells. Cluster 2 exhibited good amendment delivery in the LVZ only near the injection wells. In the PRZ at Cluster 2, amendment was effectively distributed through the full distance between injection wells. Cluster 3 exhibited the smallest amount of lateral flow and amendment was primarily distributed in the LVZ and PRZ very near the injection wells. Imaging results below the water table suggest that more amendment entered the groundwater in Cluster 2 and Cluster 3 than in Cluster 1.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Drop Analysis of the Advanced Test Reactor Fresh Fuel Shipping Container with Heavier Low-Enriched Uranium Fuel Contents

The Advanced Test Reactor Fresh Fuel Shipping Container (ATR FFSC) is a rectangular stainless steel container used for shipping radioactive material. The container is described in the ATR FFSC Safety Analysis Report (SAR). Per the ATR FFSC SAR, the ATR FFSC is designated a Type AF-96 packaging per the definition of 10 CFR §71.4, and was originally designed to transport high enriched uranium (HEU) reactor fuel elements for the Advanced Test Reactor (ATR), the Advanced Test Reactor Critical (ATRC) facility, the Massachusetts Institute of Technology Reactor (MITR), and the University of Missouri Research Reactor (MURR). The Department of Energy, National Nuclear Security Administration’s (NNSA), Office of Material Management and Minimization (M3) is working with the Idaho National Laboratory (INL) to develop and qualify new low enriched uranium (LEU) fuels and technologies for use in the ATR, ATRC, MITR, and MURR reactors. The LEU fuel elements will weigh significantly more than the current HEU designs and, combined with their associated Fuel Handling Enclosures for packaging, some configurations will exceed the 50 lbf used in the ATR FFSC qualifying drop tests. There are LEU versions of MITR, MURR, and ATR fuel elements. However, for this evaluation, drop analysis of the ATR FFSC with only the heavier ATR Low Enrichment (LOWE) fuel element is considered in this evaluation because the LOWE fuel element is the heaviest of the considered LEU fuel elements. The ATR HEU fuel element and the ATR LOWE fuel element are identical in every design aspect except for the fuel meat inside the 19 fuel plates. The LEU fuel meats are made using a U-10Mo high-density foil rather than uranium dispersed in aluminum in the HEU fuel elements. The high density of the uranium in the LEU fuel meat increases the LOWE fuel element weight to just under 44 lbf (versus the 22.1 lbf weight of the tested ATR HEU fuel element). ATR fuel elements are placed in a thin-gauge aluminum weldment called a "Fuel Handling Enclosure" during packaging. The Fuel Handling Enclosure is used to cover and protect the element during loading and unloading operations. The ATR Fuel Handling Enclosure weighs about 15 lbf per the drawings in the ATR FFSC SAR and the weight is accounted for in this evaluation. Transporting the heavier LEU fuel elements require evaluation of two issues. The first is the effect of the increased mass of the LEU fuel elements on the survivability of the ATR FFSC package following the requisite drop qualifications. The second is the effect of the increased mass of the fuel plates on the fuel element during the same drops. The ATR FFSC containing an ATR HEU fuel element in an ATR Fuel Handling Enclosure was physically dropped multiple times to qualify the container as a Type AF-96 package. The ATR FFSC SAR describes the drop tests performed with an actual ATR HEU fuel element weighing 22.1 lbf contained in a 14.3 lbf Fuel Handling Enclosure for a total payload of 36.4 lbf. Those drop tests showed that the ATR FFSC maintained containment of the ATR HEU fuel element and the fuel element was not significantly damaged. (Containment herein is not defined as a leak tight but is retention of the radioactive contents.) The purpose of the evaluation is to analytically show that, for a similar set of tests, the ATR FFSC maintains containment of the heavier ATR LOWE fuel element and to assess the damage to the fuel element during the drops. The approach was to create finite element analysis (FEA) models that produce the same results as the physical drops. Those models were then used as the benchmarks for the follow-on analyses using the heavier contents. FEA models of the drops of ATR FFSC using up to a 115 lbf fuel element were run and evaluated. Likewise, drops of a LOWE fuel element weighing 44 lbf in the ATR FFSC were run and evaluated. It is important to note that this report was done at the quality level necessary to be included in a nuclear facility safety basis. However, it is not the intent of this report to conclude the suitability of the ATR FFSC for transporting the heavier payloads. This report only describes the results of the FEA as related to the required drop scenarios. Incorporation of the FEA into the safety basis will be evaluated by the ATR FFSC design authority. The physical drop tests of the HEU fuel element and FEA drop analysis for the LOWE fuel element showed noteworthy damage to the fuel plates. An aluminum protective block was conceived to mitigate the damage. The concept requires the blocks to be placed in the fuel element between the end boxes and fuel plates. Additional FEA drop analyses were performed using the protective block. The addition of the blocks is primarily intended to mitigate the damage to the LOWE fuel element fuel plates. However, FEA drop analyses of the ATR HEU fuel element with the blocks were also performed and included for information.

42 ENGINEERING↗
DOE OSTI2021

High Flux Isotope Reactor Low-Enriched Uranium Low Density Silicide Fuel Design Parameters

High Flux Isotope Reactor (HFIR) highly enriched uranium (HEU) to low-enriched uranium (LEU) conversion activities are ongoing as part of the Department of Energy (DOE) National Nuclear Security Administration (NNSA)’s nuclear nonproliferation mission. Design activities studying the conversion of HFIR from HEU to LEU fuel explored different fuel design features and shapes with a low density uranium-silicide dispersion (U 3 Si 2 -Al) fuel, which has a uranium density of 4.8 gU/cm 3 . The goal of these studies is to generate several HFIR LEU fuel designs of varying fuel fabrication complexity that meet the current HEU performance metrics and safety requirements. The documented designs will serve as references for fuel fabrication and qualification activities. Recent advancements in modeling and simulation tools enable quick prototyping of fuel designs. Shift, a Monte Carlo neutron transport and depletion tool optimized for high-performance computing (HPC) architectures, is used for efficient fuel cycle and performance metrics calculations. The HFIR Steady State Heat Transfer Code (HSSHTC) is used to vet the thermal safety margin. Also, a new automation tool that connects all fuel design analysis steps, named Python HFIR Analysis and Measurement Engine (PHAME), has been developed to expedite the design study in an efficient and reproducible manner. Leveraging these tools, several candidate fuel designs were selected for varying fabrication complexity. This report provides design feature details for four selected HFIR LEU low density U 3 Si 2 -Al fuel designs and their corresponding performance and safety metrics. Nominal, best-estimate design parameters and irradiation conditions, including fission rate densities, power densities, heat fluxes, and cumulative fission densities are provided for candidate fuel designs relevant to framing irradiation experiments to support fuel qualification efforts. Simulations show that the low density U 3 Si 2 -Al, with design features to enhance safety, can meet HEU core performance metrics and safety requirements if the reactor power is increased from 85 MW (HEU) to 95 MW (LEU) and if the active fuel length is increased from 50.80 cm (HEU) to 55.88 cm (LEU).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Immobilization of Uranium in a Wetland 8-km Downstream from an Industrial Source (Report)

A three-day walk-over survey with portable gamma detecting equipment was conducted over a uranium-contaminated wetland located 8-km downstream from a former fuel fabrication facility. 97,000 spectra were collected, individually processed to extract uranium activity data, and then integrated with topographical information (LIDAR) to create maps. These maps reveal that 50 years after being released into the environment, 11% of the uranium is retained in the 9-ha wetland area. Combined with measurement collected closer to the source, it was determined that 94% of the released uranium is presently being held in downstream wetlands. This demonstrates that wetlands can be very effective at immobilizing contaminants for long durations. The survey also showed that previous flyover maps overestimated the size of the contaminated areas by 140% but had the advantage of being completed in orders-of-magnitude fewer labor hours. Depending on project needs, the walk-over technique may be useful to include in tandem with flyover measurements to acquire maps with greater sensitivity and spatial resolution.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

High Flux Isotope Reactor Low Enriched Uranium U-10Mo Fuel Design Parameters

Activities to convert the HFIR from HEU to LEU are ongoing as part of the US Department of Energy (DOE) National Nuclear Security Administration (NNSA) nuclear nonproliferation mission. Design activities to study the conversion of HFIR from HEU to LEU fuel explored different fuel design features and shapes with a uranium-molybdenum (U-10Mo) monolithic alloy fuel. This high-density alloy contains 90 wt % uranium and 10 wt % molybdenum and has a uranium density of 15.318gU/cm 3 . The goal of these studies is to generate several candidate HFIR LEU fuel designs of varying fuel fabrication complexity that meet the current HEU performance metrics and safety requirements. Recent advancements in modeling and simulation tools and design methods enabled a thorough analysis of the available design space with U-10Mo fuel. A surrogate model used this analysis as training data to quickly determine the performance of a design given specific design parameters. An optimization module used this surrogate model to quickly search this multidimensional search space given specific desired performance characteristics. This approach was made possible by the large available design space with U-10Mo fuel. Shift, a Monte Carlo tool optimized for high-performance computing (HPC) architectures, was used for faster calculation and better data management for reactor physics simulations. Once most of these design studies were complete, a new suite called the Python HFIR Analysis and Measurement Engine (PHAME) was developed to connect all fuel design analysis steps, making design studies more efficient and reproducible. The post-processing capabilities of these new tools are leveraged for the information provided herein. Leveraging these tools, several candidate fuel designs were selected with varying levels of feature complexity and reactor performance. This report provides design feature details for four selected HFIR LEU U-10Mo fuel designs and their corresponding performance and safety metrics. Nominal best-estimate design parameters and irradiation conditions, including fission rate densities, power densities, heat fluxes, and cumulative fission densities, are provided. Simulations show that the high uranium density of U-10Mo fuel provides a large potential design space that enables various LEU designs to meet HEU core performance metrics and safety requirements with a power increase from 85 MW (HEU) to 95 MW or 100 MW (LEU).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗