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Glasgow, David

Publications and source records attributed to Glasgow, David.

12 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

NO 2 -mediated voloxidation for iodine separation from cesium iodide surrogates

Heterogeneous NO 2 -mediated oxidation of uranium, also known as advanced voloxidation, is a proposed head-end reprocessing method for used nuclear fuel. An advantage of advanced voloxidation is the removal of volatile fission products, which complicate downstream separation and containment challenges leading to increased processing economics. Iodine, one of the volatile species of interest, has exhibited varied results in this process. Using CsI as a surrogate material, this work mimics the effect of NO 2 -based voloxidation on iodine and sheds light on the factors that influence the solid–gas phase reaction. Solid-state analysis using Fourier transform infrared attenuated total reflectance spectroscopy and scanning electron microscopy with energy-dispersive X-ray spectroscopy confirmed the conversion of CsI to CsNO 3 . Iodine separation ranged from 46% to 100% across multiple tests. Iodine separation was most effective when multiple recharges of NO 2 were administered. In conclusion, at the bench scale, liberating iodine from CsI appears to occur within 1 h, but the presence of surface H 2 O and the composition of the NO x reagent mixtures greatly influence its success.

advanced voloxidation↗
DOE OSTI2024

Iodine Capture Studies of Silver Mordenite and Novel Alternative Metal Sorbents

Radioiodine is one of the radionuclides of concern when considering reprocessing of used nuclear fuel (UNF). It is expected to be released primarily in the dissolver off-gas (DOG) stream. Sorbents that target I 2 removal from a UNF processing off gas must perform efficiently under elevated temperatures in the presence of water vapor and nitrogen oxide gasses (NOx). Many studies have been conducted that examine the adsorption of iodine species in the presence of NOx gases through reduced silver-exchanged mordenite (Ag 0 Z), a zeolite mineral. This study characterizes the effects of iodine, water vapor, and nitrogen dioxide on the adsorption of I 2 by Ag 0 Z at lower temperatures suitable for comparison with novel Cu- and Bi- bearing sorbents which offer a non-RCRA alternative. When exposed to ‘harsh’ conditions such as I 2 , H 2 O, and NO 2 vapors carried by air, Cu 2 S-polyacrylonitrile (PAN) is more efficient than Ag 0 Z and all other novel sorbents, providing with 11% I capture efficiency. However, under ‘ideal’ conditions in which I 2 is carried by dry air), Ag 0 Z is more efficient than Cu 2 S-PAN and other novel sorbents, providing 74% I capture efficiency. To investigate this result and the overall performance of all sorbents tested, several characterization techniques were employed, including thermal gravimetric analysis, scanning electron microscopy – electron dispersion X-Ray spectroscopy, powder X-ray diffraction, and X-ray photoelectron spectroscopy. These techniques are discussed within this report. This document also includes a design and test plan update from Idaho National Laboratory as the next step in testing the high performing sorbents from thin-bed tests carried out by Oak Ridge National Laboratory.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Modeling and Design of Micro Flux Monitors for Neutron Fluence Measurements

Measuring neutron flux is critical to understanding the characteristics of nuclear reactors and can provide information on the energy spectrum, total neutron fluence, and operational history of a facility. Such measurements are typically performed by irradiating high-purity materials in an operating reactor for a known length of time. By measuring and identifying activation products in the irradiated materials and referencing the energy-dependent cross section of each isotope, the neutron energy spectrum and total fluence can be determined. Historical techniques use a variety of small, individually pressed foils or wires for dosimeter devices, which can result in excessively high levels of radioactivity and complications in handling after irradiation. The present work describes a new neutron dosimetry concept, named the micro flux monitor, that uses semiconductor device fabrication processes to deposit thin films of flux monitoring materials in precise geometries onto a SiC wafer. This approach allows multiple dosimetry materials to be located on a single device, simplifying irradiation and subsequent analysis. The geometry, and therefore mass, of each material can be custom-tailored for a specific application—considering the expected neutron flux, duration of irradiation, and device retrieval time of a facility. Additionally, hundreds of identical dosimetry monitors can be produced simultaneously, reducing fabrication costs. Finally, the extremely thin metal dimensions of these devices could improve the fidelity of neutron fluence measurements by reducing neutron self-shielding and gamma self-attenuation effects.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Realistic application of short-lived fission product delayed neutron, gamma-ray analysis for simultaneous nondestructive trace quantification of U, Pu mixtures on cellulose swipes

Detection and characterization of fissile traces are of interest to the international nuclear nonproliferation community, including the International Atomic Energy Agency. Pre-inspection check samples are analyzed by neutron activation analysis at the High Flux Isotope Reactor operated by the Oak Ridge National Laboratory under the umbrella of the IAEA Network of Analytical Laboratories. The simultaneous quantification of U and Pu mixtures was accomplished using the combined delayed neutron (DN) delayed gamma-ray (DG) method to analyze cellulose swipes with actinide loading <1ng in a blind field trial. The total fissile quantity was measured by the DN counts and the relative proportions of U, Pu were determined by calibration of the 104 Tc / 141 Ba fission product count ratio using known mixtures. The DNDG method demonstrated high accuracy in flagging the presence of 239 Pu in uranium down to <100 pg mass loading. In conclusion, peak significance tests helped to control false positive Pu flagging and simultaneous quantification of U and Pu loading was accomplished on samples that passed the significance tests.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Cumulative fission yields of short-lived fission products from 235 U and 239 Pu measured by HPGe gamma-ray spectroscopy

In this study, we present a preliminary investigation focused on determining cumulative fission yields for short-lived fission products. Our analysis involves examining gamma spectra from the irradiated samples of 235 U and 239 Pu using the High Flux Isotope Reactor. The motivation stems from the observed discrepancy in the antineutrino energy spectrum within the range of 5 to 7 MeV. While several hypotheses have been proposed, a thorough analysis of fission yields provides an additional way of gaining insight into this unexplained phenomenon. Our study suggests that the measured gamma rays from 100 Nb, 140 Cs and 95 Sr are consistent with the expected values. However, 93 Rb, 96 Y, 97 Y and 142 Cs cannot be quantified due to insufficient statistics, interference from other gamma rays and the Compton scattering background. Additionally, the calculated cumulative fission yields based on the measured 140 Cs and 95 Sr are found to be consistent with the JEFF3.3 fission yield library. In conclusion, the present work shows that the potential of improving gamma-ray spectroscopy in the fission yields as a means to improve our understanding of the antineutrino spectrum.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Short-lived noble gas effluent trends from a research reactor

An understanding of anthropogenic sources of radioactive noble gases in the atmosphere is needed to enhance the discrimination ability of the International Monitoring System's sensors. These sources include commercial and research nuclear reactors and medical isotope production facilities. While abiding by local environmental ordinances these facilities all emit noble gas radioisotopes through normal operation. Here, this research presents measurements and analysis of noble gas isotopes ( 41 Ar, 135 Xe, 135m Xe, 137 Xe, 138 Xe, 87 Kr, 88 Kr, and 89 Kr) made directly at the stack of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. The Xe and Kr noble gases are concurrently observed with 41 Ar, a neutron activation product, when the reactor is operational. The magnitude of the Xe and Kr noble gases released is not constant over the HFIR cycle, but they temporally match the 41 Ar trend. An isotope activity ratio analysis of these shorter lived isotopes combined with the observation of the cycle's temporal trend helps understand the noble gas production mechanism at the HFIR. Isotopes with short half-lives are not useful for long-range environmental monitoring. However, these measurements could potentially be combined with atmospheric modeling to predict the background source term of the longer-lived Xe ratios at a monitoring station.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Volume 4: Detection Systems and Ultra-Cold Neutrons

As part of the Sustaining and Enhancing Nuclear Science (SENSe) initiative at Oak Ridge National Laboratory (ORNL), the prospect of adding new detection systems to support High Flux Isotope Reactor (HFIR) operations and to advance scientific research has prompted many ideas and discussions regarding potential features, configurations, locations, and applications. A working group of ORNL staff members was organized to further develop the concepts and recommending one or more configurations to best support future HFIR operations and scientific capacities and to provide order-of-magnitude cost estimates and timing. The areas of investigation included fast access detection systems, non-scattering beamline instruments, shielded detection instruments, and an ultracold neutron source. In each case, the focus was to develop world leading capabilities that would be unmatched by any other facility.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Measurement of the Effective Capture Cross Section of 238 Np in the High Flux Isotope Reactor

The cross sections of 237 Np and 238 Np are important for accurate modeling and simulation of 238 Pu in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Uncertainties in these cross sections can impact the ability to predict and optimize the target design and loading for 238 Pu production targets. The effective capture cross section of 237 Np in the location of pneumatic tube 1 in HFIR was measured as a first step in the measurement of the 238 Np capture and fission cross sections. Here, we describe the flux measurements, 237 Np experiments, and data analysis of the 237 Np capture cross section in HFIR.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

An approach to separating Pu, U, and Ti from high-purity graphite for isotopic analysis by MC-ICP-MS

Information about elemental and isotopic systematics of ultra-trace level actinides (e.g. U and Pu) and main group elements (e.g. Ti) present within nuclear grade graphite is vital to the nuclear community for improved reactor operation and security. In support of this, extensive effort has been placed on improving analysis methods (i.e., inductively coupled plasma-mass spectrometry). However, significantly less effort has been devoted to the optimization of chemical separation methods. Within the separation community, commercially available Eichrom™ resins are often employed, as their elution characteristics for various elements have been well studied, but the direct optimization of actinides and trace metal separations from a single sample have not been widely investigated. Here, methods using various Eichrom pre-packed cartridges were explored to achieve separation of ultra-trace levels of U, Pu, and Ti from a variety of graphite samples. Once the validity of the combined separation scheme was established using certified reference materials, the method was applied to historic, unirradiated and irradiated, graphite samples. For all samples investigated, precise isotope ratio measurements for the titanium isotope systems were made.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗