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Abrecht, David G.

Publications and source records attributed to Abrecht, David G..

8 records
All sourcesNASA NTRSDOE OSTI
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 OSTI2023

Ab initio modeling and thermodynamics of hydrated plutonium oxalates

An ab initio study on the plutonium oxalate hydrates: Pu 2 (C 2 O 4 ) 3 ·10H 2 O and Pu(C 2 O 4 ) 2 ·6H 2 O, using PBE exchange-correlation with D3 dispersion correction and Hubbard correction for the plutonium atoms, was performed and compared to experimental vibrational spectral and thermodynamic property values. Here we demonstrated that this technique can accurately predict the experimental infrared spectra Pu(III) oxalate hydrate, as well as the Raman peak of PuO 2 (used to calculate the thermodynamic properties of the oxalates). For Pu(IV) oxalate hydrate we found that our predicted structure agreed qualitatively with PXRD measurements, the only available experimental determination of the structure. Using this method at standard temperature and pressure, we predicted standard enthalpies of formation of -6,755 kJ mol -1 and -3,923 kJ mol -1 and standard Gibbs free energy of formation of -5,899 kJ mol -1 and -3,386 kJ mol -1 for Pu 2 (C 2 O 4 ) 3 ·10H 2 O and Pu(C 2 O 4 ) 2 ·6H 2 O, respectively.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Characterizing PuO 2 Powder at Oak Ridge National Laboratory for the Multi-Lab Plutonium Process Signatures Campaign FY 2022

This report highlights the development of advanced analytical capabilities at Oak Ridge National Laboratory to characterize PuO 2 powder to support the Multi-Lab Plutonium Process Signatures Campaign in FY 2022. Two samples from a batch of 76 statistical precipitation runs were packaged and shipped from Pacific Northwest National Laboratory to Oak Ridge National Laboratory at the beginning of FY 2022. Techniques were used to characterize the sample, including powder X-ray diffraction, Raman spectroscopy, and diffuse reflectance spectroscopy. A brief description of each technique, the results and discussion, and future work are described in this report. Opportunities to expand current capabilities and include additional analytical capabilities for the next campaign will also be discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Engineering Evaluation of Barium Buildup in a Decayed CsCl Sealed Source and Potential Impact for Cesium Release from a Breached Source

For more than 50 years, radioactive 137 Cs has been a major source material for radioactive sealed sources, usually constructed as cesium chloride (CsCl) salt loaded into double-walled, stainless-steel capsules. A complication develops as 137 Cs decays to 137 Ba since this process creates a strongly reducing environment inside the capsule. A potential hazard exists if the capsule is breached and air ingress induces rapid exothermic oxidation, and this mechanism is suspected to be responsible for the well-known contamination incident at the Harbor View facility in Washington state. For this study, many thermodynamic evaluations were performed to assess the internal state of capsules after several decades of decay and to describe the potential oxidation if the capsule contents were to suddenly be exposed to air. Results suggest that reduction of impurities such as Cu, Fe, Pb, and Cr to metal will occur. If these impurities are lacking, it is possible that even Ba metal will form. In most cases, rapid oxidation can occur, and the exothermic reactions are sufficient to vaporize portions of the contents, which would include the remaining 137 Cs.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Modeling Induction Stirring and Particle Tracking in Molten Uranium

Two independent numerical models have been developed to simulate the behavior of carbon impurities in molten uranium metal. Informed by experimental parameters, one model was created using the commercial software Star-CCM+ and compared with another developed using open-source codes, including OpenFOAM, Finite Element Method Magnetics (FEMM) and a First Passage Kinetic Monte Carlo (FPKMC) approach. The target experimental system features a 404g uranium metal charge containing an average carbon concentration of 139 ppm which was melted in a vacuum induction furnace at 1400° C then resolidified. The microstructures of the uranium and its impurities before and after melting have been characterized and reported separately. Prior to simulating the uranium-carbon system described, the numerical models were validated using a previously published nonradioactive experimental system to ensure agreement with expected output values. Focus has been placed on modeling velocity fields under induction stirring and impurity particle trajectories.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Solvent Exchange

This report summarizes the work of evaluating the chemical stability of phosphoramidic acid, N-N-diethyl-, bis(2-ethylhexyl) ester under the Solvent Exchange LDRD project, as a potential solvent exchange chemistry and analog to solvent exchange chemistries used in the TALSPEAK and ALSEP processes as alternatives to the use of tri-butyl phosphate in the PUREX process.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Improving Ductility of Hydride Embrittled Zirconium (Final Report)

One of the main limiting factors on the lifetimes of nuclear fuel rod claddings is the formation of hydrides and associated detrimental effects on mechanical properties. In this study, we examined the behavior of hydrides in zirconium and zircaloy-4 after tensile stress is applied to determine whether the more ductile γ-ZrH phase was stabilized by the mechanical stress. We did not find a significant increase in the ratio of γ-ZrH phase to δ-ZrH1.5 phases after tensile stress is applied in previously hydrided metals. Previous reports indicate that this stabilization does occur when zircaloy is stressed and hydrided simultaneously, indicating that the formation of the γ-ZrH phase may reverse upon relaxation or may require the stress in-situ during the hydriding process to form in significant quantities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗