Engineering topics

Mayes, Richard

Publications and source records attributed to Mayes, Richard.

6 records

DOE OSTI2024

Integrated Off-gas System: A Preconceptual Design of an Integrated Off-Gas Treatment System

The U.S. has a vested interest in the advancement of nuclear energy to achieve aggressive net-zero goals, with reprocessing and recycling of used nuclear fuels (UNF) playing a vital role. It will not be possible to meet U.S. regulatory requirements without robust off-gas treatment, so it is crucial to advance treatment technologies to facilitate the design of future reprocessing facilities. For many years, teams of researchers across the U.S. Department of Energy (DOE) National Laboratory complex have been investigating off-gas treatment technologies for the capture and removal of volatile radionuclides (i.e., 85 Kr, Xe, 14 C, and 129 I) and oxides of nitrogen (NO X ) that are produced from reprocessing. These investigations have been focused on developing individual technologies for the capture of Kr, Xe, iodine, and CO 2 . Capture technologies for each constituent were tested independently from one another by utilizing nonradioactive surrogates to simulate simplified off-gas streams. The tests have been relatively small, laboratory-scale experiments of up to approximately 1 L/minute total gas flow rate. To increase the readiness of these technologies for deployment, an integrated test system with a larger-scale capacity is needed to bridge the gap between promising bench scale and fully scalable UNF reprocessing off-gas treatment. This document contains the goals, design basis, functional requirements, preconceptual design, and cost estimates for an integrated off-gas demonstration system for the capture and removal of NO x , Kr, Xe, CO 2 , and iodine at 10× higher throughput than earlier laboratory studies. The order-of-magnitude cost estimate for this system is approximately $\$$886,000. Next phases include conceptual design, detailed design, fabrication, and commissioning.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Behavior of contaminants and simulant fission products in the advanced chlorination process

The University of Tennessee, Knoxville and Oak Ridge National Laboratory have completed proof-of principle testing on an advanced sulfur-based zirconium and aluminum chlorination process for the recovery of Zr or Al from spent fuel claddings and purification of the products. This process could be used to recover hafnium-free nuclear-grade zirconium for reuse or disposal as low-level waste. Testing showed that even non-optimized reaction protocols could produce zirconium product streams from simulated irradiated fuel with purities >90%. The chlorination process was tested on representative fission product species that would be found in Advanced Test Reactor fuel (ATR). First, fission product generation was determined through modeling and simulation of the ATR. The generation of fission products was used to selected contaminants and potentially volatile fission product simulants for laboratory testing. The goal is to identify which species may react with the sulfur-chlorine reagents and be transported into the advanced chlorination off-gas.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Strategies toward the Synthesis of Advanced Functional Sorbent Performance for Uranium Uptake from Seawater

A polymer fiber-based adsorbent (AF1) composed of acrylonitrile and itaconic acid functional groups was synthesized by a radiation-induced graft polymerization technique onto hollow-gear shaped polyethylene (PE) fibers. Investigation of the optimum reaction parameters for the conversion of grafted cyano moieties into amidoxime groups was conducted by the reaction with hydroxylamine at different temperatures and periods in a variety of aqueous and organic solvents. The 13 C CP/MAS spectra of amidoximated AF1 adsorbent fibers from water–methanol and dimethyl sulfoxide (DMSO) revealed that both the cyclic imide dioxime and open-chain amidoxime were formed through the reaction. The conversion from amidoxime to imide dioxime was found to occur slowly and gradually with the increasing reaction time. The quicker diffusion of DMSO as compared to that of water–methanol, in the grafted trunk PE fiber, resulted in faster kinetics of the amidoximation reaction. The uranium adsorption capacity of the amidoximated AF1 samples was determined after (i) 24 h contact with sodium-based brine spiked with uranium and (ii) 56 days of exposure to filtered seawater (Sequim Bay, WA, USA) in flow-through columns. Finally, the uranium extraction performance of the adsorbents after exposure to filtered seawater was consistent with the laboratory screening results, and the amidoximated AF1 samples (in DMSO at 70 °C for 3 h) exhibited the highest 56 day uranium adsorption capacity (5.04 ± 0.15 g U/kg-ads) with faster adsorption kinetics compared to the pristine AF1 adsorbents.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2021

Insight into the Solid Electrolyte Interphase Formation in Bis(fluorosulfonyl)Imide Based Ionic Liquid Electrolytes

The formation of the solid electrolyte interphase (SEI) in an ionic liquid electrolyte of 0.5 m lithium bis(fluorosulfonyl)imide (LiFSI) in 1-ethyl-3-methylimidazolium bis(fluorosulfonyl)imide at high cell voltages (1.7–1.9 V) is investigated in ordered mesoporous carbon (OMC) based Li metal cells using an operando small-angle neutron scattering (SANS) technique coupled with electrochemical impedance spectroscopy and ex situ X-ray photoelectron spectroscopy (XPS). It is demonstrated that discharging the OMC Li metal cells to ≈2 V and holding the cell voltage constant induces a rapid current increase with time, confirming extensive reduction and SEI formation. XPS analysis reveals that LiF is formed at open cell voltage (OCV), which is attributed to the carbenes generated at the lithium negative electrode because of its reaction with EMIm cation diffusing to and initiating the reaction with FSI - anions at the carbon positive electrode. It is confirmed that the chemical reaction at OCV and electrochemical reduction at high cell voltage of the FSI - anion plays a protective role against EMIm cation co-intercalation into the carbon positive electrode during the initial discharge. Operando SANS studies also suggest that slight differences occur in the surface composition and reaction mechanism as a function of cell voltage.

1-ethyl-3- methylimidazolium bis(fluorosulfonyl)im↗

DOE OSTI2020

Understanding Materials Characteristics of Palladium Lanthanide Cermet Wires as a Mimic of Californium Wire Production

Californium-252 (Cf-252) has been synthesized on a large scale (mg) since the 1970s for use as a portable neutron source. With a half-life of 2.645 years and a spontaneous fission fraction of 3.09%, Cf-252 emits 2.31 × 10 12 neutrons/gram second. Today most of the Cf-252 produced in the world is made at Oak Ridge National Laboratory (ORNL) using the High Flux Isotope Reactor (HFIR) and the Radiochemical Engineering Development Center (REDC). Because of its highly radioactive nature, all manipulations of Cf-252 need to be performed in hot cell facilities by highly trained personnel. The Cf-252 produced at ORNL is done at the REDC, which is conveniently located next to the HFIR. Cf-252 is sold as bulk wire, which has been described as palladium (Pd) metal incasing californium oxide (Cf 2 O 3 ). The process was developed by Mosley et al. at Savannah River Laboratory (SRL) where it was modeled after the thorium work done by Fuschillo et al. This process has been used since the 1970s to synthesize this unique wire, but its fundamental chemistry and materials characteristics were never openly reported. Here we aim to better understand the chemistry and the materials characteristics of Cf-252 wire production for process improvements and to better our fundamental understanding of the mid-late actinides. Though there are many challenges in working with radioactive elements such as Cf, the isotope Cf-252 is even more complicated due to its neutron emission. One challenge with neutron emission is the ability of neutrons to penetrate much deeper through materials, causing radiation damage. This makes using instrumentation to characterize the chemistry and materials properties of this process difficult. For this reason, this study will first focus on terbium (Tb) as a surrogate for Cf. Currently Tb is used as a carrier or filler in Cf production when small amounts of Cf are being processed. This, along with their similar size and charge, makes Tb a reasonable surrogate for this study.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Thermal and radiation response of 4H–SiC Schottky diodes with direct-write electrical contacts

In this work, a high-sensitivity 4H–SiC temperature sensor and an alpha detector have been fabricated using additively printed metal contacts. The surface morphology and electrical conductivity of the printed electrodes were established prior to Schottky diode development. 4H–SiC Schottky diodes with direct-write printed silver contacts on the 5 μ m-thick epilayer on 4H–SiC were characterized electrically in terms of the forward and reverse current–voltage and high-frequency capacitance–voltage characteristics. Furthermore, the turn-on voltage of the Schottky diodes, as established from the forward current–voltage characteristics measured up to a temperature of 400 °C, showed a linear temperature dependence. Schottky diodes with direct-write printed Ag electrodes were able to measure alpha particles emitted from Americium-241. The high temperature and radiation response of the Schottky diodes show their suitability for multi-modal sensor fusion on the 4H–SiC platform for harsh environment applications.

42 ENGINEERING↗