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At least 595 records · Page 33
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

U.S. HIGH PERFORMANCE RESEARCH REACTOR PRELIMINARY DESIGN MILESTONE FOR CONVERSION TO LOW ENRICHED URANIUM FUEL

The U.S. reactors that regularly refuel with highly enriched uranium (HEU) have undertaken conversion to low-enriched uranium (LEU) using a high-density alloy of uranium-10 wt% molybdenum (U-10Mo) that is under development for the U.S. High Performance Research Reactors (USHPRR). The U.S. National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Reactor Conversion Program is pursuing USHPRR conversion. The USHPRR consists of nine national laboratories and three universities working with a uranium processing plant and commercial fuel fabricator to convert five reactors and an associated critical assembly. Feasibility designs were completed in 2009 for all USHPRR. Since then, the USHPRR have completed key milestones leading to preliminary designs. Between 2007 and 2011, the Fuel Qualification (FQ) Pillar conducted seven irradiation campaigns consisting of 14 large size plates and over 60 mini plates with U-10Mo fuel with a zirconium interlayer. The Preliminary Report on U-Mo Monolithic Fuel for Research Reactors submitted for review to the U.S. Nuclear Regulatory Commission (NRC) in 2017 summarizes these irradiation tests. The Fuel Fabrication (FF) Pillar has installed and is operating a fabrication pilot line at BWX Technologies, Inc. In tandem, the Reactor Conversion (RC) Pillar, has been working with each USHPRR to mature conversion designs and to include recent data. As of 2018, all five USHPRR have completed LEU designs. This includes three Preliminary Safety Analysis Reports (PSARs) submitted between 2014 and 2017 to the NRC by the National Institute of Standards and Technology, University of Missouri, and Massachusetts Institute of Technology for their reactors NBSR, MURR, and MITR, respectively. In 2018, the two USHPRR reactors regulated by the U.S. Department of Energy (DOE) at Idaho National Laboratory (ATR) and Oak Ridge National Laboratory (HFIR) completed optimized conceptual or preliminary designs, with ATR confirming a ‘base fuel’ design of the same fabrication complexity as required for MURR, MITR and NBSR. Instead of available U-10Mo designs, HFIR is pursuing a longer silicide (U3Si2) fuel design with a fuel meat density of 4.8 gU/cm3. However, HFIR conversion will require additional fuel qualification due to the complex fuel design containing boron and since HFIR exceeds existing silicide qualification limits. These preliminary designs allow for a commercial fabrication demonstration, and subsequent irradiation qualification of these fuel element designs in order to complete final SARs that allow for USHPRR conversion to LEU fuel.

Wilson, E. H.↗
DOE OSTI2019

Method to produce uranium silicides

The method described herein may be characterized as reacting uranium dioxide with carbon to produce uranium carbide, and, reacting the uranium carbide with a silane, a silicon halide, a siloxane, or combinations thereof, and excess hydrogen to produce uranium silicide.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Update on the High Precision Titration Method for Uranium Assay Supported by NBL Program Office

The NBL Program Office (NBL PO) has coordinated with Oak Ridge National Laboratory (ORNL) to implement the High Precision Titration (HPT) method for uranium assay measurements. The measurement method has been successfully developed and qualified, therefore ORNL has progressed to performing analyzes critical to the mission of the NBL PO. The HPT method is vital to producing the next generation of certified reference materials for uranium assay and isotopic abundance. The work described here focused on establishing traceability of the method to the NIST SRM 136 potassium dichromate series, and focused on investigating small differences between SRM 136e and 136f. The NBL PO evaluated the history of all of its primary uranium reference materials, including CRM 112A natural uranium metal, and worked with NIST in investigating the small differences in the dichromate SRM’s. ORNL performed comparative experiments on SRM 136e and 136f using CRM 112A. The results of the experiment will be presented here, along with NBL PO’s plans to re-evaluate the CRM 112A certified uncertainty to comply with JCGM 100, “Guide to the expression of uncertainty in measurement.”

Rogers, Kayron↗
DOE OSTI2024

Commercialization of High-Density High Assay Low Enriched Uranium Fuel Systems

The Office of Reactor Conversion and Uranium Supply (NA 231) at the National Nuclear Security Administration leads the conversion effort for the United States High Performance Research Reactors (USHPRR). These reactors are the final civilian reactors in the US to transition from High Enriched Uranium (HEU) to high assay low enriched uranium (HALEU). Each of these reactors represents unique capabilities and no currently available fuel system meets their needs for conversion. The Fuel Fabrication (FF) Pillar of the USHPRR project is responsible for the fabrication of experimental elements, conversion elements, and establishing a commercial economical production capability. FF is also responsible to share with the domestic and international community the theoretical knowledge gained. Other pillars within the USHPRR project provide the experimental and conversion fuel designs, assist the reactors with licensing activities, and ensure the entire fuel cycle is evaluated. Over the last decade, FF has worked with the production partners at Y-12 National Security Complex (Y-12) and BWXT Nuclear Operations Group, Research and Test Reactors (BWXT). Y-12 has begun processing the alloy feedstock for the conversion elements with a qualified process. BWXT has started the final fabrication of the experimental elements. Once the experimental elements are complete, BWXT will begin conversion element fabrication. The FF Pillar resides at Pacific Northwest National Laboratory (PNNL) and uses PNNL, universities, commercial vendors, and the DOE national laboratory system to evaluate process development activities to improve the process steps. FF supports the fabrication of two high density fuel systems, monolithic U-10Mo (Figure 1) and Uranium Silicide (Figure 2). The U-10Mo fuel system is further along the development process. FF assists in long term planning with the production partners. This includes ramping production of the elements from experimental quantities to annual steady state needs. As part of the ramp up, opportunities to improve yield and product quality are identified to ensure the fuel systems are cost effective.

Catalan, Michael A. [BATTELLE (PACIFIC NW LAB)]↗
DOE OSTI2026

Impact of Molten Gallium on the Microstructure and Corrosion Behavior of Aluminum and Uranium-Aluminum Alloys for Used Nuclear Fuel Reprocessing

Test reactors around the world utilize highly enriched uranium fuel to achieve high neutron fluxes for materials testing. Once spent, the remaining uranium is a valuable resource for subsequent fuel fabrication. However, some of these test reactor cores consist of curved plate-type fuel elements, fabricated using aluminum alloy 6061 (AA6061) cladding to encapsulate a uranium-aluminum alloy (UAlx) fuel matrix. These assemblies require non-standard reprocessing approaches for uranium recovery, as aluminum dissolves readily in acidic solutions, generating large volumes of waste and complicating downstream chemical separations. In this work, we investigate a novel chemical decladding strategy based on the interaction between AA6061/UAlx and molten gallium (Ga). Ga is known to induce severe degradation of aluminum metal through liquid metal embrittlement (LME), even at relatively low Ga concentrations. By penetrating the aluminum crystal lattice, Ga disrupts grain cohesion and facilitates fracture or dissolution of the aluminum matrix. Thermodynamic analysis of the Al–Ga binary phase diagram suggests that Ga may offer a viable pathway to selectively weaken or dissolve the AA6061 cladding, and potentially the aluminum component of the UAlx fuel matrix within. To this end, parametric experiments were performed at 50 °C and 100 °C across a range of Al–Ga atomic fractions. At lower Al fractions, the AA6061 was completely molten after 2 hours of exposure to the Ga metal. In contrast, samples with higher Al fractions (0.9 Al, 0.1 Ga) contained residual solids after 2 hours, which were characterized by microstructural examination using electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). These Al-Ga compositions were also evaluated using FactSage thermodynamic modeling to further elucidate the relationship between phase diagram behavior and LME.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Obtaining Diluent for High-Enrichment Uranium Hexafluoride in Two- and Three-Stage Cascades

The production of diluent for high-enrichment uranium in a cascade from waste uranium hexafluoride with unknown {sup 234}U content is examined. To secure high-quality diluent and reduce separative work consumption, it is proposed that the diluent be produced in two stages with sorting and extraction of the best intermediate produced in terms of {sup 234}U concentration after the first stage. The worst intermediate products are best used at the second stage to obtain low-enrichment uranium hexafluoride satisfying ASTM C 996-15 specifications. In addition, the obtained new wastes can be used in a third stage to produce diluent with low {sup 234}U concentration.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Nuclear forensic analysis on uranium ore concentrates: A multi-laboratory intercomparison study

A collaboration between the United States and Japan performed nuclear forensic analyses on a set of 7 uranium ore concentrate (UOC) samples with known origins to improve methodologies and best practices in the field. Overall, good agreement between the 4 participating laboratories was achieved for uranium assay, uranium isotope composition, strontium isotope composition, and trace element analysis. In conclusion, the combination of these four signatures is sufficient to distinguish the different origins among these studied UOCs.

and nuclear chemistry↗
DOE OSTI2023

Computational studies of impurity migration during induction stirring of molten uranium

Understanding and controlling impurity behavior in actinide metal casting processes are foundational for efficient part production yet remain major challenges for researchers and industry. To help provide insight regarding impurity distribution during actinide metal casting, we have developed computational fluid dynamics (CFD) models for a laboratory-scale system using commercial and open-source code. Multiple simulation frameworks allow for improved confidence in the resulting outputs, while taking advantage of the maturity and convenience offered by commercial platforms and simultaneously maintaining the transparency and flexibility often provided by open-source software. Here, we describe multiple experiment-informed models designed to simulate a specific laboratory system in which uranium melt, containing a known starting concentration of carbon impurity, is electromagnetically stirred in an induction furnace. Here, the goal of the simulations is to predict the motion of uranium carbide microparticles in the velocity field of the melt. Prior to simulating the uranium-carbon system, numerical models were validated using a previously published nonradioactive experimental system. Effects of the size and shape of impurity particles in the models were investigated and agree with experimental findings. Simulation of smaller particles (< 50 µm) shows more homogenous distribution throughout the stirred melt. With increased particle size (100 µm), the body forces, which include the buoyancy force, dominate over the drag force, causing larger particles to move toward the crucible walls and upward in the system.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Reactive transport of uranium in a groundwater bioreduction study: Insights from high-temporal resolution 238 U/ 235 U data

In this paper, we conducted a detailed investigation of U isotopes in conjunction with a broad geochemical investigation during field-scale biostimulation and desorption experiments. This investigation was carried out in the uranium-contaminated alluvial aquifer of the Rifle field research site. In this well-characterized setting, a more comprehensive understanding of U isotope geochemistry is possible. Our results indicate that U isotope fractionation is consistently observed across multiple experiments at the Rifle site. Microbially-mediated reduction is suggested to account for most or all of the observed fractionation as abiotic reduction has been demonstrated to impart much smaller, often near-zero, isotopic fractionation or isotopic fractionation in the opposite direction. Data from some time intervals are consistent with a simple model for transport and U(VI) reduction, where the fractionation factor (ε = +0.65‰ to +0.85‰) is consistent with experimental studies. However, during other time intervals the observed patterns in our data indicate the importance of other processes in governing U concentrations and 238 U/ 235 U ratios. For instance, we demonstrate that departures from Rayleigh behavior in groundwater systems arise from the presence of adsorbed species. We also show that isotope data are sensitive to the onset of oxidation after biostimulation ends, even in the case where reduction continues to remove contaminant uranium downstream. Finally, our study and the described conceptual model support the use of 238 U/ 235 U ratios as a tool for evaluating the efficacy of biostimulation and potentially other remedial strategies employed at Rifle and other uranium-contaminated sites.

58 GEOSCIENCES↗
DOE OSTI2015

Defect structures induced by high-energy displacement cascades in γ uranium

Displacement cascade simulations were conducted for the c uranium system based on molecular dynamics. A recently developed modified embedded atom method (MEAM) potential was employed to replicate the atomic interactions while an embedded atom method (EAM) potential was adopted to help characterize the defect structures induced by the displacement cascades. The atomic displacement process was studied by providing primary knock-on atoms (PKAs) with kinetic energies from 1 keV to 50 keV. The influence of the PKA incident direction was examined. The defect structures were analyzed after the systems were fully relaxed. The states of the self-interstitial atoms (SIAs) were categorized into various types of dumbbells, the crowdion, and the octahedral interstitial. The voids were determined to have a polyhedral shape with {110} facets. The size distribution of the voids was also obtained. The results of this study not only expand the knowledge of the microstructural evolution in irradiated c uranium, but also provide valuable references for the radiation-induced defects in uranium alloy fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Uranium(VI) coordination polymers with pyromellitate ligand: Unique 1D channel structures and diverse fluorescence

Three new coordination polymers of uranium(VI) with pyromellitic acid (H{sub 4}btca) have been synthesized and structurally characterized. (ED)[(UO{sub 2})(btca)]·(DMSO)·3H{sub 2}O (1) (ED=ethylenediammonium; DMSO=dimethylsulfoxide) has a lamellar structure with intercalation of ED and DMSO. (NH{sub 4}){sub 2}[(UO{sub 2}){sub 6}O{sub 2}(OH){sub 6}(btca)]·~6H{sub 2}O (2) has a 3D framework built from 7-fold coordinated uranyl trinuclear units and btca ligands with 1D diamond-shaped channels (~8.5 Å×~8.6 Å). [(UO{sub 2}){sub 2}(H{sub 2}O)(btca)]·4H{sub 2}O (3) has a 3D network constructed by two types of 7-fold coordinated uranium polyhedron. The unique μ{sub 5}-coordination mode of btca in 3 enables the formation of 1D olive-shaped large channels (~4.5 Å×~19 Å). Vibrational modes, thermal stabilities and fluorescence properties have been investigated. - Graphical abstract: Table of content: three new uranium(VI) coordination polymers with pyromellitic acid (H{sub 4}btca) have been synthesized via room temperature and hydrothermal synthesis methods, and structurally characterized. Two to three dimensional (3D) frameworks are revealed. All 3D frameworks have unique 1D large channels. Their vibrational modes, thermal stabilities and photoluminescence properties have been investigated. - Highlights: • Three new coordination polymers of U(VI) with pyromellitic acid (H{sub 4}btca). • Structures from a 2D layer to 3D frameworks with unique 1D channels. • Unusual µ{sub 5}-(η{sub 1}:η{sub 2}:η{sub 1}:η{sub 2:}η{sub 1}) coordination mode of btca ligand. • Vibrational modes, thermal stabilities and luminescent properties reported.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2013

First-principles DFT+U modeling of defect behaviors in anti-ferromagnetic uranium mononitride

A series of point defects in uranium mononitride (UN) have been studied by first-principles DFT+U calculations. The influence of intrinsic defects on the properties of UN was explored by considering the anti-ferromagnetic (AFM) order along the [001] direction. Our results show that all the point defects lead to obvious volume swelling of UN crystal. Energetically, the interstitial nitrogen defect is the most favorable one among single-point defects in UN crystal with the formation energy of 4.539 eV, while the N-Frenkel pair becomes the most preferable one among double-point defects. The AFM order induces obvious electron spin polarization of uranium towards neighboring uranium atoms with opposite spin orientations in UN crystal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Evaluation of a measurement system for Uranium electrodeposition control to radiopharmaceuticals production

For 2016, studies by international bodies forecast a crisis in the supply of Molybdenum ({sup 99}Mo), which is the generator of {sup 99m}Tc, widely used for medical diagnoses and treatments. As a result, many countries are making efforts to prevent this crisis. Brazil is developing the Brazilian Multipurpose Reactor (RMB) project, under the responsibility of the National Nuclear Energy Commission (CNEN). The RMB is a nuclear reactor for research and production of radioisotopes used in the production of radiopharmaceuticals and radioactive sources, broadly used in industrial and research areas in Brazil. Electrodeposition of uranium is a common practice to create samples for alpha spectrometry and this methodology may be an alternative way to produce targets of low enriched uranium (LEU) to fabricate radiopharmaceuticals, as {sup 99}Mo, used for cancer diagnosis. To study the electrodeposition, a solution of 10 mM uranyl nitrate, in 2-propanol, containing uranium enriched to 2.4% in {sup 235}U, with pH = 1, was prepared and measurements with an alpha spectrometer were performed. These studies are justified by the need to produce {sup 99}Mo since, despite using molybdenum in bulk, Brazil is totally dependent on its import. In this project, we intend to obtain a process that may be technologically feasible to control the radiation targets for {sup 99}Mo production. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2019

Integration of Local Knowledge is a Key for the Remediation Success of Legacy Uranium Mine Sites - 19564

A key to success, during all aspects of the remediation process on legacy uranium sites, and easily transferable between countries and projects, is the integration of local knowledge. However, before this knowledge can be integrated, local trust and project understanding must be developed. Local trust and project understanding can only be developed through a principle-centred approach to a shared project vision, two-way communication of knowledge, alignment of project goals, and understanding the local needs early in the remediation process. Once public confidence is obtained, local trust and project understanding must be maintained throughout the entirety of the remediation process. Project understanding is an ongoing endeavor that needs continuous communication and varied methods to succeed. During the management of assessment, remediation, and post remediation monitoring on the Saskatchewan Research Council's CLEANS project (remediation of 37 legacy uranium mine sites in Northern Saskatchewan, Canada), integration of local knowledge has been an instrumental tool attributing to success. During site assessment, integration of local knowledge has included the development and implementation of land use surveys to determine realistic travel and occupancy on and adjacent to these legacy sites, and site specific traditional knowledge and land use studies that have contributed to the development of a detailed knowledge of local land use, resource utilization, country food intake, and human history of the sites. The local knowledge gained during the assessment process is invaluable as it provides information required to support remediation planning, such as: valued ecosystem components, human trophic utilization, length of site occupancy, cultural value assessment, and future land use considerations. During remediation planning, local and traditional knowledge can also be incorporated through determination of plant species suitability for revegetation, development and selection of remediation options, and setting realistic sustainable development targets for local project inclusion. During site remediation operations, local involvement can be utilized through direct labour and equipment, cultural resource monitoring (i.e. archaeological and heritage protection), local logistics and historic information transfer, and stakeholder communications and engagement support. Post remediation, local involvement is an efficient mechanism for long-term monitoring, and care and maintenance activities. Although there may be differences between countries, regions and sites, the remediation process on legacy uranium mines can universally be more successful through the integration of local knowledge; while developing rights/stakeholder engagement and capacity building. (authors)

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

Nuclear Fuel Based on Uranium-Zirconium Carbonitride

The results of Investigations of the nitrogen and uranium pressure in the system U–Zr–C–N in the temperature range 1873–2073 K and the constructed section of the U–Zr–C–N phase diagram at 1873 K are presented. The behavior of the solid fission products in carbonitride fuel, the interaction of the fuel with a pyrolytic carbon coating, and the behavior of micro fuel elements based on such fuel during high temperature firing and in the presence of a temperature gradient are examined. The conditions for the preparation of fuel compositions of the type uranium carbonitride in a carbide-silicon matrix and some of its physicochemical and mechanical properties are determined. It is noted that there is promise in using fuel compositions and micro fuel elements based on uranium-zirconium carbonitride for high-temperature applications.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Reactivity Runaway Reduction When Using Enriched Uranium in a Lead Fast Reactor

The possibilities for reactivity runaway reduction in a fast reactor when using enriched uranium are examined. Specifically, an approach supposing partial compensation of reactivity runaway by the addition of Np or Am, as consumable absorbers, into the starting load is analyzed. It is shown that from the standpoint of the development of large-scale nuclear power generation based on BREST-type reactors the use of a uranium starting load becomes inexpedient in implementing this approach. A starting load consisting of a mixture of enriched uranium and plutonium from spent RBMK fuel is studied. The results indicate that it is possible to develop a core with reactivity runaway not exceeding the effective delayed-neutron fraction during the entire period of operation. The neutronics characteristics were calculated using the Arktika multigroup diffusion code. The RISK code was used to determine the change in the nuclide composition.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

A Uranium Bioremediation Reactive Transport Benchmark

A reactive transport benchmark problem set has been developed based on in situ uranium bio-immobilization experiments that have been performed at a former uranium mill tailings site in Rifle, Colorado, USA. Acetate-amended groundwater stimulates indigenous microorganisms to catalyze the reduction of U(VI) to a sparingly soluble U(IV) mineral. The interplay between the flow, acetate loading periods and rates, microbially-mediated and geochemical reactions leads to dynamic behavior in metal- and sulfate-reducing bacteria, pH, alkalinity, and reactive mineral surfaces. The benchmark is based on an 8.5 m long one-dimensional model domain with constant saturated flow and uniform porosity. The 159-day simulation introduces acetate and bromide through the upgradient boundary in 14-day and 85-day pulses separated by a 10 day interruption. Acetate loading is tripled during the second pulse, which is followed by a 50 day recovery period. Terminal electron accepting processes for goethite, phyllosilicate Fe(III), U(VI), and sulfate are modeled using Monod-type rate laws. Major ion geochemistry modeled includes mineral reactions, as well as aqueous and surface complexation reactions for UO2++, Fe++, and H+. In addition to the dynamics imparted by the transport of the acetate pulses, U(VI) behavior involves the interplay between bioreduction, which is dependent on acetate availability, and speciation-controlled surface complexation, which is dependent on pH, alkalinity and available surface complexation sites. The general difficulty of this benchmark is the large number of reactions (74), multiple rate law formulations, a multisite uranium surface complexation model, and the strong interdependency and sensitivity of the reaction processes. Results are presented for three simulators: HYDROGEOCHEM, PHT3D, and PHREEQC.

bioremediation↗