Search NASA⌕ Search

Engineering topics

O'Hara, Matthew J.

Publications and source records attributed to O'Hara, Matthew J..

14 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Radiation Modeling of Z1: Validation of a Novel Radioisotope System

This paper presents the development of a radiation model for a novel strontium-90 (Sr-90) radioisotope heat source developed by Zeno Power Systems (Zeno), which demonstrates a groundbreaking fuel and shielding design that significantly reduces weight compared to traditional concepts. A Monte Carlo N-Particle (MCNP) model has been created to assess the effectiveness of this fuel and shielding design, however validation of the MCNP model is desired. Zeno has developed a prototype device (Z1) to aid in this model validation effort. This project is a collaborative effort between Zeno, the University of Dayton Research Institute (UDRI) and Pacific Northwest National Laboratory (PNNL), where the Z1 prototype was constructed and characterized.

RTG↗
DOE OSTI2024

Revival and Technology Transfer of 90SrTiO3 Production for Heat Source Applications

Demand for radioisotope power systems (RPS) appears to be on the rise, and it is unlikely that this new demand can be met by plutonium-238. As a result, new heat source designs are under development. One such design is Z1. Z1 is a strontium-90 demonstration heat source developed by Zeno Power Systems in collaboration with the University of Dayton Research Institute and the Pacific Northwest National Laboratory. SrTiO3 was chosen as the fuel form for Z1. SrTiO3 was used to produce dozens of terrestrial RPS decades ago. While this indicates the technology is proven, the skill and experience necessary to effectively produce SrTiO3 has been lost. Recapturing SrTiO3 production technology therefore became necessary. Development started with non-radioactive surrogate experiments using natural strontium. This work showed that the old Oak Ridge flowsheets for SrTiO3 production are quite robust, but some critical pieces of the technology are left out of the old documentation. Specifically, particle sizes of the reagents are critical to producing a phase pure SrTiO3. If the reagent particles are not the right size, then a significant Sr3Ti2O7 phase impurity is observed. This phase impurity was proven to be the result of an incomplete reaction between the strontium and titanium precursors. This technology was transferred to Pacific Northwest National Laboratory where the technology was adapted for use in a hot cell. With the adaptations complete, this technology was used to produce the fuel for Z1 late last year. This represents the first new heat source design produced in the U.S. in over 40 years.

90Sr, strontium titanate, SrTi03, RTG, Radioisotop↗
DOE OSTI2024

Systems and methods for separating radium from lead, bismuth, and thorium

Methods for separating Ra from Pb, Bi, and Th are provided, the methods can include: providing a first mixture comprising Ra, Pb, Bi, and/or Th; providing a system that can include: a first vessel housing a first media; a second vessel in fluid communication with the first vessel, the second vessel housing a second media; and a third vessel in fluid communication with the second vessel, the third vessel housing a third media; and exposing the first mixture to the first media within the first vessel then, through the fluid communication, exposing the first remainder to the second media in the second vessel, then, through fluid communication, exposing the next remainder to the third media in the third vessel, the exposing separating the Th and Bi from the Ra and Pb, and the Ra from the Pb. Methods for separating Ra from being associated with a media are also provided. The methods can include: exposing the Ra and media to a chelating agent to form a mixture comprising the Ra complexed with the chelating agent.

O'Hara, Matthew J.↗
DOE OSTI2023

System and process for purification of Astatine-211 from target materials

A new column-based purification system and approach are described for rapid separation and purification of the alpha-emitting therapeutic radioisotope 211 At from dissolved cyclotron targets that provide highly reproducible product results with excellent 211 At species distributions and high antibody labeling yields compared with prior art manual extraction results of the prior art that can be expected to enable enhanced production of purified 211 At isotope products suitable for therapeutic medical applications such as treatment of cancer in human patients.

O'Hara, Matthew J.↗
DOE OSTI2022

Demonstration of low-level biogenic fuel content using quench curve and direct liquid scintillation counting (LSC) methods

In this work, an investigation into the direct method for liquid scintillation counting (LSC) quantification of biogenic fuel content in fuel blendstock is presented. This method development intentionally used a colored matrix (~99 wt%), fossil diesel fraction, and low-level biogenic fractions (~1 wt%) to determine the applicability of the LSC technique in colored blend mixtures. LSC procedures for quench correction, including the use of an instrument internal standard (Quantulus SQP values), color quench curves, and an internal 14 C standard spike on samples containing low-level amounts of biogenic gasoline, jet, and diesel mixed with fossil fuel, were compared. Samples were analyzed on both Perkin-Elmer Quantulus and Tri-Carb instruments, and the 14 C contents were compared with those determined by accelerator mass spectrometry (AMS). Results from the Tri-Carb instrument showed that percent differences of <10 % compared to AMS-reported values are achievable at 5-hour count times despite colored samples. A comparison between the impact of chemical and color quenching in the fuel samples showed that color quenching in highly colored samples reduces efficiency significantly more than chemical quenching, indicating that chemical quench curves are not appropriate for highly colored biofuel/fossil-fuel blended samples. Results indicate that both the direct method with internal spike quench correction and the use of a color-quench curve provide accurate results for 1 % biogenic fuel blends. Additionally, we have explored the use of Fourier transform infrared (FTIR) spectroscopy in measuring the biogenic content of the colored blended fuel samples. Preliminary results show the presence of biogenic carbon-derived fingerprints that are absent in fossil-derived sample. However, further work is needed to develop an FTIR-based quantitative method.

09 BIOMASS FUELS↗
DOE OSTI2022

Affordable Realtime Viral Detection: Demonstrated with SARS COV-2

The near real-time detection of airborne particles-of-interest is needed for avoiding current/future threats. The incorporation of imprinted particles into a micelle-based electrochemical cell produced a signal when brought into contact with particle analytes (such as SARS-COV-2), previously imprinted onto the structure. Nanoamp scales of signals were generated from what may’ve been individual virus-micelle interactions. The system showed selectivity when tested against similar size and morphology particles. The technology was compatible with airborne aerosol sampling techniques. Overall, the application of imprinted micelle technology could provide near real-time detection methods to a host of possible analytes of interest in the field.

47 OTHER INSTRUMENTATION↗
DOE OSTI2022

Corrosion of U233-Doped Uranium Oxide using Microfluidics Methods

The aim of work this year has been to investigate the role of alpha (α)-radiation and hydrogen (H 2 ) on the corrosion of uranium oxide (UO 2 ) using a microfluidic device. The microfluidic device, termed the Particle-Attached Microfluidic Electrochemical Cell, (PAMEC), enables monitoring of the UO 2 electrochemical corrosion potential (E corr ) under different environments, including de-aerated conditions and in the presence of dissolved H 2 . The Si 3 N 4 window allows us to study morphological and chemical changes under an electron microscope. We have previously demonstrated that the PAMEC matches the results from bulk electrochemical tests with UO 2 [1]. We used the high specific activity uranium (U) isotope, 233 U, (t 1/2 = 160,000 years) incorporated into UO 2 , to generate a localized a-field. The objective of the experiments were to mimic the radiation environment that would be experience at the surface of aged spent nuclear fuel (SNF) during long-term geologic disposal under anoxic conditions. Wittman et al. [2] predicted that in the presence of a pure a-radiation field and under H 2 conditions, the concentration of the radiolytic oxidant H 2 O 2 would be suppressed or even eliminated. In a UO 2 corrosion experiment this would be exhibited through a lowering of the measured UO 2 corrosion potential compared to identical conditions in the absence of dissolved H 2 . We found that the predictions of Wittman and co-workers were supported and that the corrosion potential of the 233 U-doped UO 2 in solution lowered with presence of H 2 gas and increased in the absence of H 2 , when under anoxic conditions. The PAMEC experiments indicate that H 2 O 2 has been eliminated in a solution sparged with H 2 while exposed to an a-radiation field. The corrosion potential of the a-doped 233 U(10%)- 238 UO 2 in a solution sparged with Ar/H 2 matched the corrosion potential of 238 UO 2 in solution sparged with air. This clearly demonstrated that the H 2 O 2 had been eliminated and that the only oxidant present was O 2 in this system in complete agreement with the modeling results of Wittman et al. These results point to the need to improve the Fuel Matrix Degradation (FMD) Process Model training data set that is being used in the FMD surrogate model that is being developed for the repository program. The incorporation of realistic radiation chemistry will improve the scientific basis for the FMD Model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Anion exchange and extraction chromatography tandem column isolation of zirconium-89 ( 89 Zr) from cyclotron bombarded targets using an automated fluidic platform

The long-lived positron emitter 89 Zr is a highly promising nuclide employed in diagnostic Positron Emission Tomography (PET) imaging. Methods of radiochemical processing to obtain 89 Zr for clinical use are traditionally performed with a single hydroxamate resin column. In this work, we present a tandem column purification method for the preparation of high-purity 89 Zr from cyclotron bombarded Y metal foils. The primary column is a macroporous, strongly basic anion exchange resin on styrene divinylbenzene co-polymer. The secondary microcolumn, with an internal volume of 33 µL, is packed with an extraction chromatography resin (ExCR) loaded with HDEHP. A condition of “inverted selectivity” is presented, wherein the 89 Zr elution from the primary column is synonymous with the load condition on the secondary column. The ability to transfer 89 Zr from one column to the next allows two sequential column clean-up methods to be performed prior to the final elution of the 89 Zr product. This approach assures delivery of very high purity 89 Zr . The tandem column purification process has been implemented into a prototype automated fluidic system. Optimization of the method is presented, followed by evaluation of the process using seven cyclotron bombarded Y metal foil targets. Once optimized, we found that 93.7±2.3% of the 89 Zr present in the foils was recovered in the secondary column elution fraction (0.8 M oxalic acid). Radiochromatograms of the product elution peaks enabled determination of full width at half-maximum (FWHM) and 89 Zr collection yields as a function of volume. Because of the small size of the secondary microcolumn, a 89 Zr product volume of ~0.28 mL is reported, which provides a substantially increased nuclide concentration over traditional methods. Finally, we evaluated the transchelation of the resulting 89 Zr oxalate product to deferoxamine mesylate (DFOM) salt. We observed effective specific activities (ESA) and bindable metals concentrations ([M B ]) that exceed those reported by the traditional single hydroxamate column method.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2022

System, emanation generator, and process for production of high-purity therapeutic radioisotopes

An isotope production system, emanation generator, and process are disclosed for production of high-purity radioisotopes. In one implementation example, high-purity Pb-212 and/or Bi-212 isotopes are produced suitable for therapeutic applications. In one embodiment the process includes transporting gaseous radon-220 from a radium-224 bearing generator which provides gas-phase separation of the Rn-220 from the Ra-224 in the generator. Subsequent decay of the captured Rn-220 accumulates high-purity Pb-212 and/or Bi-212 isotopes suitable for direct therapeutic applications. Other high-purity product isotopes may also be prepared.

O'Hara, Matthew J.↗
DOE OSTI2022

Determination of Biogenic Content by 14 C in fuels using Liquid Scintillation Analysis Laboratory Analytical Procedure (Laboratory Analytical Procedure)

This procedure covers the determination of the percent 14 C in biogenic gasoline, jet fuel, and diesel range hydrocarbons prepared from distillation of hydrotreated pyrolysis oils and blended with fossil derived hydrocarbons (e.g. fossil diesel, gasoline, jet, or toluene). Either an internal 14C standard or a color quench curve is used to correct for potential signal quenching effects. Recent publications provide some background of direct LSC methods used for determination of biogenic fuel content.

02 PETROLEUM↗
DOE OSTI2022

System and process for purification of astatine-211 from target materials

A new column-based purification system and approach are described for rapid separation and purification of the alpha-emitting therapeutic radioisotope 211At from dissolved cyclotron targets that provide highly reproducible product results with excellent 211At species distributions and high antibody labeling yields compared with prior art manual extraction results of the prior art that can be expected to enable enhanced production of purified 211At isotope products suitable for therapeutic medical applications such as treatment of cancer in human patients.

O'Hara, Matthew J.↗
DOE OSTI2022

Separations of U/Pu and Np/Pu using fluoride volatility

Here the importance of uranium/plutonium and neptunium/plutonium separations at large scale or at environmental levels is such that the reactivity of nitrogen trifluoride (NF 3 ) with respect to other fluorinating reagents warranted investigation. We show that NF 3 does not convert PuO 2 or PuF 4 to PuF 6 , during separations of U/Pu and Np/Pu, even under forcing conditions. Herein, this is analytically demonstrated to trace level. This rather exclusive behavior has been reported for only one other reagent: BrF 3 . Confidence in this result now allows for rapid and selective separations of volatile UF 6 and NpF 6 from plutonium. Preliminary data concerning the separation of 233 Pa from 237 Np is also described.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗