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At least 1,675 records · Page 93
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Design and Performance Testing of a Linear Array of Position-Sensitive Virtual Frisch-Grid CdZnTe Detectors for Uranium Enrichment Measurements

Abstract— Arrays of position-sensitive virtual Frisch-grid CdZnTe (CZT) detectors with enhanced energy resolution have been proposed for spectroscopy and imaging of gamma-ray sources in different applications. The flexibility of the array design, which can employ CZT crystals with thicknesses up to several centimeters in the direction of electron drift, allows for integration into different kinds of field-portable instruments. These can include small hand-held devices, compact gamma cameras and large field-of-view imaging systems. In this work, we present results for a small linear array of such detectors optimized for the low-energy region, 50-400 keV gamma-rays, which is principally intended for incorporation into hand-held instruments. There are many potential application areas for such instruments, including uranium enrichment measurements, storage monitoring, dosimetry and other safeguards-related tasks that can benefit from compactness and isotope-identification capability. The array described here provides a relatively large area with a minimum number of readout channels, which potentially allows the developers to avoid using an ASIC-based electronic readout by substituting it with hybrid preamplifiers followed by digitizers. The array prototype consists of six (5x5.7x25 mm3) CZT detectors positioned in a line facing the source to achieve a maximum exposure area (~10 cm2). Each detector is furnished with 5 mm-wide charge-sensing pads placed near the anode. The pad signals are converted into X-Y coordinates for each interaction event, which are combined with the cathode signals (for determining the Z coordinates) to give 3D positional information for all interaction points. This information is used to correct the response non-uniformity caused by material inhomogeneity, which therefore allows the usage of standard-grade (unselected) CZT crystals, while achieving high-resolution spectroscopic performance for the instrument. In this presentation we describe the design of the array, the results from detailed laboratory tests, and preliminary results from measurements taken during a field test.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Design and Performance Testing of a Linear Array of Position-Sensitive Virtual Frisch-Grid CdZnTe Detectors for Uranium Enrichment Measurements

Arrays of position-sensitive virtual Frisch-grid CdZnTe (CZT) detectors with enhanced energy resolution have been proposed for spectroscopy and imaging of gamma-ray sources in different applications. The flexibility of the array design, which can employ CZT crystals with thicknesses up to several centimeters in the direction of electron drift, allows for integration into different kinds of field-portable instruments. These can include small hand-held devices, compact gamma cameras and large field-of-view imaging systems. In this work, we present results for a small linear array of such detectors optimized for the low-energy region, 50-400 keV gamma-rays, which is principally intended for incorporation into hand-held instruments. There are many potential application areas for such instruments, including uranium enrichment measurements, storage monitoring, dosimetry and other safeguards-related tasks that can benefit from compactness and isotope-identification capability. The array described here provides a relatively large area with a minimum number of readout channels, which potentially allows the developers to avoid using an ASIC-based electronic readout by substituting it with hybrid preamplifiers followed by digitizers. The array prototype consists of six (5x5.7x25 mm3) CZT detectors positioned in a line facing the source to achieve a maximum exposure area (~10 cm2). Each detector is furnished with 5 mm-wide charge-sensing pads placed near the anode. The pad signals are converted into X-Y coordinates for each interaction event, which are combined with the cathode signals (for determining the Z coordinates) to give 3D positional information for all interaction points. This information is used to correct the response non-uniformity caused by material inhomogeneity, which therefore allows the usage of standard-grade (unselected) CZT crystals, while achieving high-resolution spectroscopic performance for the instrument. In this presentation we describe the design of the array, the results from detailed laboratory tests, and preliminary results from measurements taken during a field test.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Neutronic and Thermal-Hydraulic Feasibility Studies for High Flux Isotope Reactor Conversion to Low-Enriched Uranium U3Si2-Al Fuel

Oak Ridge National Laboratory (ORNL) operates the High Flux Isotope Reactor (HFIR) for the US Department of Energy Office of Science to perform neutron scattering experiments, produce isotopes, and conduct materials irradiation research. ORNL is funded by the National Nuclear Security Administration (NNSA) Office of Materials Management and Minimization to evaluate the conversion of HFIR’s fuel from high-enriched uranium (HEU) to low-enriched (LEU). Due to potential fabrication issues with the complex UMo fuel design required to convert HFIR, NNSA requested ORNL to evaluate an alternate LEU fuel form – U3Si2-Al dispersion fuel. Neutronic and thermal-hydraulic feasibility analyses were performed with Shift and HSSHTC, respectively, to predict reactor performance and thermal safety margins. A number of designs were proposed and evaluated using an iterative approach in an effort to show that reactor performance could match that obtained using HEU fuel and that thermal safety margins were adequate. This study concludes that conversion of HFIR with U3Si2-Al LEU (19.75 wt.%) fuel with 4.8 gU/cm3 is feasible if, among other requirements, the fuel meat region is centered and symmetric about the fuel plate thickness centerline, the active fuel zone length is increased 5.08 cm to 55.88 cm, the proposed fabrication tolerances can be met, and the fuel can be qualified for HFIR conditions.

Chandler, David↗
DOE OSTI2018

Action levels for airborne natural uranium in the workplace: Chemical and radiological assessments

A method is described for deriving two levels of action—an investigation level (IL) and an immediate action level (IAL)—for different forms and mixtures of the natural uranium (U) isotopes 234U, 235U, and 238U in air in the workplace. An IL indicates the need to confirm the validity of moderately elevated measurements of airborne U and adequacy of confinement controls and determine whether work limitations are appropriate. An IAL indicates that safeguards should be put into place immediately, including removal of workers from further exposure until conditions are acceptable. Derivations of ILs and IALs are based on latest radiation protection guidance, information on chemical toxicity of U, and biokinetic models for U. An action level (IL or IAL) is the more restrictive of two derived values, the action level based on U as a chemical hazard and the action level based on U as a radiation hazard.

Leggett, Richard Wayne↗
DOE OSTI2019

Shining a light on amorphous U 2 O 7 : A computational approach to understanding amorphous uranium materials

Mixed-phase and low-symmetry systems are widely observed among uranium oxide materials. Amorphous-U 2 O 7 forms during the calcination of studtite, a uranyl peroxide mineral. Using a genetic algorithm search for stable crystalline phases, we have identified a potentially stable phase of U2O7 that shares structural features with experimentally observed amorphous U2O7 samples. The crystalline structure is expected to undergo a solid–solid phase change around 12 GPa.

Shields, Ashley E.↗
DOE OSTI2019

Assessment of Low Enriched Uranium Fueled Core Configurations for the Versatile Test Reactor

The Versatile Test Reactor (VTR) Program was initiated in FY-2017. This multi-laboratory effort supported by the U.S. Department of Energy has the long-term objective of completing a Versatile Test Reactor concept with associated cost and schedule estimate that enables a fully informed DOE decision within 3 years, to have a fully operating VTR by 2026 if acquisition is authorized, and to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. The VTR concept uses sodium as coolant, features a core power of 300 MWth, and employs reactor-grade plutonium as fissile material in a U-Pu-10Zr fuel alloy. The reference core layout provides 10 assembly locations for high-flux irradiation testing with peak fast flux in excess of 4x1015 n/cm2-s. In order to establish a comparison point with other types of fuel, alternative core configurations were investigated using Low Enriched Uranium (LEU) fuel. The objective of this study was to modify the reference VTR core such as to use LEU fuel and assess the performance tradeoff, while retaining the same core envelope.

Stauff, Nicolas E.↗
DOE OSTI2019

Influence of Calcite on Uranium(VI) Sorption onto Montmorillonite Clay

Currently, most proposed methods for the long-term storage of nuclear waste include engineered buffer systems (EBS) in underground geologic repositories. The design of EBS typically involves a layer of geological buffer material such as bentonite. Bentonite is an ideal material for these systems, since it largely consists of the swelling smectite clay montmorillonite, which provides two types of surface sites available for the sorption of radionuclides, such as uranium(VI) (U(VI)). However, the presence of calcite, a common accessory mineral in bentonite, may potentially alter the sorption of U(VI) onto montmorillonite in multiple ways. First, calcite provides an additional surface for U(VI) sorption reactions. In addition, calcite dissolution will lead to dissolved calcium in porewater solutions, which may in turn affect U(VI) solution speciation, U(VI) ion exchange reactions, and the stacking and interlayer spacing of smectite layers. Last, a potential formation of U(VI)-Ca-carbonate co-precipitates may remove U(VI) from solution. As a first step, we characterized the sorption of U(VI) onto montmorillonite in the presence and absence of dissolved calcium at various ionic strength conditions in batch sorption envelope experiments. The presence of dissolved calcium resulted in decreased U(VI) sorption onto montmorillonite in acidic environments, but in no apparent changes in U(VI) sorption in near-neutral pH ranges. At higher pH values, there may be a possible decrease in U(VI) sorption, which needs to be further investigated. The observed trends at low pH can be explained by the preferential binding of Ca 2+ cations relative to UO 2 2+ , which resulted in a competition between the two cations for clay surface sites. This hypothesis is further supported by the results of additional sorption experiments, which were performed at various ionic strengths, but in the absence of calcium. In a next step, we will expand our experiments to include mixtures of calcite and montmorillonite as solids.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Effects of Calcite Impurities in Bentonite Buffer Systems on Uranium(VI) Solution Speciation and Sorption Behavior

In many nuclear waste disposal scenarios, bentonite is the proposed buffer material surrounding nuclear waste containers. Its main mineralogical component, Na-montmorillonite clay, is expected to limit radionuclide mobility to slow diffusion processes due to the clay’s low porosity and permeability, and the high sorption affinity of clay surface complexation and cation exchange sites for many radionuclides. Given its mineral structure, montmorillonite provides two types of porosities that can be accessible to radioactive contaminants: macropores between clay particles and interlayer spaces within clay particles, which become dominant at increasing degrees of bentonite compaction. Despite the fact that bentonite naturally contains a variety of mineral impurities (e.g., quartz, feldspars and calcite), most previous studies have only focused on radionuclide interactions with pure Na-montmorillonite phases. Hence, in this study, we will investigate the influence of calcite impurities on uranium(VI) (U(VI)) solution speciation and sorption behavior in bentonite systems. In the presence of water, calcite impurities are expected to lead to milli-molar concentrations of dissolved calcium over a wide range of pH conditions. Consequently, Ca 2+ could – at least partially – replace Na + on clay cation exchange sites, with further implications for exchange reactions of cationic U(VI) species. In addition, ternary U(VI)-Ca-carbonate solution species (CaUO 2 (CO 3 ) 3 2- and Ca 2 UO 2 (CO 3 ) 3 ) will largely replace U(VI)-carbonate species (UO 2 (CO 3 ) 3 4- and UO 2 (CO 3 ) 2 2- ) at alkaline pH. These larger, but lower-charged U(VI) species, may show different sorption and diffusion characteristics, e.g. if their access to negatively-charged clay interlayer spaces is different from binary U(VI)-carbonate complexes. In this presentation, we will first demonstrate the impacts of calcite impurities on U(VI) solution speciation based on PHREEQC simulations. We will then discuss first results from U(VI) sorption studies in these systems. This research is being performed using funding received from the DOE Office of Nuclear Energy's Nuclear Energy University Program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Uranium(VI) Sorption onto Montmorillonite Clay in the Presence of Ca2+/Calcite Impurities

The storage of nuclear waste creates a series of significant environmental risks. One attempt to minimize these potential impacts is the use of underground geological repositories, where waste containers are surrounded by geologic buffer materials creating an engineered barrier. One buffer material that has been of interest is bentonite, which is largely composed of montmorillonite clay. These barriers limit the mobility of radioactive contaminants, such as uranium(VI), into environmental systems based on contaminant sorption onto the clay material. System conditions, such as pH, the presence of mineral impurities such as calcite, and the concentrations of U(VI) and atmospheric CO2 also affect the sorption behavior of U(VI) onto these clay barriers. This study focuses on the effects of Ca2+ and calcite (CaCO3), a common impurity found in montmorillonite on the sorption behavior of U(VI). Computer simulations of the chemical solution species (chemical form) of U(VI), in the absence and presence of Ca2+/calcite impurities are performed in parallel with laboratory sorption experiments. Experimental data will be compared to modeling results to help obtain further insights regarding the impacts of Ca2+/calcite impurities on U(VI) solution speciation and sorption behavior onto montmorillonite.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Relevant Uranium Corrosion II

This presentation gives a chronology of relevant uranium oxidation with the plethora of experimental units converted to film thickness in nanometers and time in mega-seconds.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Nuclear Deactivation and Downgrade of Enriched Uranium Facilities at the Y-12 National Security Complex

Buildings 9206 and 9212 at the Y-12 National Security Complex are both Hazard Category 2 nuclear facilities from the Manhattan Project era that supported various missions over the decades, including the processing of enriched uranium. The production mission for Building 9206 ended in the 1990s, while Building 9212 is still an active production facility. The disposition process for contaminated excess facilities is generally described in the U.S. Department of Energy (DOE) Guide DOE O 430.1A, Life Cycle Asset Management, which provides guidance on facility surveillance and maintenance, deactivation, decommissioning, and transition in supplemental guides DOE G 430.1-2 through 430.1-5. Historically, production facilities similar to Buildings 9206 and 9212 have not always been cleaned out and deactivated prior to being excessed, which significantly complicates the decommissioning phase. Many legacy facilities were shut down before the formal DOE guidance was established. This paper presents the practical application and implementation of the DOE guidance for the nuclear deactivation of Buildings 9206 and 9212, and summarizes the lessons learned for application across the DOE enterprise.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

URANIUM HOLDUP MONITORING WITH COMPTON IMAGING AS FUNCTION OF DEPTH AND MASS

Mass quantification using standoff gamma detection is a logical progression for MC&A applications in uranium processing facilities. While obtaining a ground truth from physical collection and weighing of masses is ideal and routine, it poses a logistical problem from both a standpoint of ALARA programs as well as interruptions to operations. The standard approach is the Generalized Geometry in Holdup (GGH) method that can apply for many cases where geometries of the holdup match simulated models; however, there are many cases where a model may not exist or a feature of the holdup geometry is such that it does not match a model particularly well.1 Gamma imaging at standoff distances and outside process equipment has potential to address these shortcomings and improve mass estimations while reducing personnel exposure; however, algorithms still require a significant degree of benchmarking to build sufficient confidence in algorithms before displacing GGH in some procedures. This paper presents the creation of a set of laboratory standards of depleted U3O8 in fixed size containers covering a range of masses and effective thicknesses up to the infinite thickness for the 185-keV line. These blocks are identically sized aluminum boxes loaded such that a thin, flat face is placed towards the imager, while the back side contains closed cell foam to apply pressure to keep the powder in a fixed position at the front face as the sources are manipulated for measurements. Included is an analysis of the mass estimation using a commercially available Compton imaging instrument where multiple geometries are created showing limitations in current methodologies.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Nuclear Deactivation and Downgrade of Enriched Uranium Facilities at the Y-12 National Security Complex

Buildings 9206 and 9212 at the Y-12 National Security Complex are both Hazard Category 2 nuclear facilities from the Manhattan Project era that supported various missions over the decades, including the processing of enriched uranium. The production mission for Building 9206 ended in the 1990s, while Building 9212 is still an active production facility. The disposition process for contaminated excess facilities is generally described in the U.S. Department of Energy (DOE) Guide DOE O 430.1A, Life Cycle Asset Management, which provides guidance on facility surveillance and maintenance, deactivation, decommissioning, and transition in supplemental guides DOE G 430.1-2 through 430.1-5. Historically, production facilities similar to Buildings 9206 and 9212 have not always been cleaned out and deactivated prior to being excessed, which significantly complicates the decommissioning phase. Many legacy facilities were shut down before the formal DOE guidance was established. This paper presents the practical application and implementation of the DOE guidance for the nuclear deactivation of Buildings 9206 and 9212, and summarizes the lessons learned for application across the DOE enterprise.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗