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Bruffey, Stephanie

Publications and source records attributed to Bruffey, Stephanie.

7 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

A novel protocol to recycle zirconium from zirconium alloy cladding from used nuclear fuel rods

A potential new zirconium recycling protocol has been demonstrated using unirradiated Zircaloy-based claddings and depleted uranium oxide for the chemical removal of zirconium alloy cladding in used nuclear fuel (UNF) rods from light water reactors. This protocol is based on the application of three new scientific findings. First, a new lower temperature chlorination reaction of zirconium with sulfur monochloride is described. Second, the high solubility of zirconium chloride in thionyl chloride is used to separate it from uranium oxide fuel and fission and activation contaminants. Finally, in the third step, the zirconium chloride is purified by simple recrystallization from thionyl chloride. Utilizing sulfur chloride solvents for a lower temperature liquid-based chemical digestion and purification of zirconium alloy claddings reduces technical complications experienced by current high temperature gas phase chlorination strategies, such as contamination of product streams. This novel protocol has been demonstrated on a 3–50 g scale of unirradiated zirconium alloy, with no significant change in the time required for complete chlorination (3–4 h). This zirconium chlorination protocol has also been performed in the presence of depleted uranium oxide pellets. The depleted fuel pellets do not affect zirconium chlorination, and the chlorination protocol does not chemically or physically alter the fuel pellets. Finally, a preliminary description of an industrial protocol to recycle nuclear grade zirconium from UNF rods is presented.

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

Initial Assessment of TeF 6 Adsorption

Research and development that supports the management of off-gases from nuclear fuel reprocessing has historically been focused on the off-gas streams that arise from aqueous reprocessing technology. However, as Gen-IV reactor development pathways move toward deployment, alternative spent nuclear fuel (SNF) processing and disposition pathways have been considered more actively. This work is focused upon aspects of fluoride volatility (FV) processing. The versatility of this method for deployment against multiple types of spent fuel encourages the continued advancement of both the primary separations processes and the secondary processes, including waste treatment, material control and accountability, and engineering designs. Recent work noted that TeF 6 , the most highly volatile fluorinated compound produced during FV processing, did not have a clear abatement technology recommended in the literature. Thus, this work performed scoping tests on activated alumina and copper shot to assess whether they could be used to remove TeF 6 from gas streams that bear F 2 . Previous work on this topic was not well described in the literature and was not performed with excess F 2 in the stream as would be typical of spent fuel processing via FV. To support an understanding of the concentration of TeF 6 contacting the adsorbent beds, a series of preliminary testing identified the TeF 6 production rate and the equilibrium concentration of TeF 6 in the gas stream contacting the adsorbent. Excess F 2 was determined to not affect the ability of activated alumina to remove TeF 6 from the gas stream quickly and completely. A determination of whether excess F 2 affected the distribution depth of Te in the sorbent bed is still pending analysis of the used sorbent. Literature suggests that when activated alumina is near saturation, TeF 6 could migrate from the sorbent bed. Future testing should investigate this possibility. Copper metal did not adsorb TeF 6 in the presence of F 2 across the sorbent temperature range of 50 to 335°C. F 2 was fully removed by the copper bed. Although preliminary thermodynamics would indicate adsorption to be energetically favorable, other factors that impact adsorption (e.g., slow kinetics, excess fluorine on the copper surface, an unfavorable transition state) are likely preventing the adsorption of TeF 6 at an easily measurable rate. The work also investigated TeF 6 production from multiple forms of Te in the temperature range of 100 to 250°C. No previous study had assessed the initial reaction rates for this process. A carefully designed study allowed determination of the activation energy for the production of TeF 6 from Te metal. The substantial amount of data collected during this study merits analysis beyond what is described here. A more in-depth kinetic analysis will be pursued. Additional analytical results will provide the ability to benchmark adsorption coefficients for TeF 6 , understand the distribution of TeF 6 within the alumina bed, and better understand the effect of F 2 partial pressure on Te fluorination. The data from this report, as supplemented by these additional analyses, will be submitted to a peer-reviewed journal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Revisions of ORNL 188 W Process Based on Nonradiological Experiments

Tungsten-188 is in widespread use in the 188 W(t 1/2 = 69 d)/ 188 Re(t 1/2 =16.9 h) biomedical generator. Oak Ridge National Laboratory (ORNL) has been providing this product to the world since 1999. At ORNL, 188 W is produced via irradiation in ORNL’s High Flux Isotope Reactor (HFIR). Enriched 186 W targets in the form of sintered metallic pellets or rings achieve a compact loading in the irradiation vessel, providing a high yield per unit target. The enrichment of the target is >90% 186 W and this isotope undergoes double neutron capture to produce the desired 188W product. While 188 W is produced by neutron bombardment, 191 Os(t 1/2 = 15.4 d) is simultaneously produced as a byproduct requiring separation from 188 W by post irradiation treatment. In the current processing pathway, the irradiated W metal pellets or rings are first converted into an oxide form of WO 3 by heating the irradiated W metal target at 750°C in a quartz reaction vessel inside a vertical furnace under a constant flow of air. During heating, W metal reacts with oxygen in the air to produce WO 3 , which is soluble in 6 M NaOH for later purification process. This oxidation process also converts 188 Os (the decay daughter of 188 W) and 191 Os (the irradiation produced byproduct) into OsO 4 , a highly volatile and toxic gas. The gaseous effluents driven from the quartz reaction vessel are passed through a scrubbing array to remove OsO 4 before the air is discharged from the process. This heterogeneous oxidation method simultaneously achieves two goals: 1). converting metal target to a soluble oxide form and 2). separating volatile OsO 4 away from the solid WO 3 product, although harmful 191 OsO 4 is unfavorable but being taken care of by the down road scrubbing array. In the past twenty years the existing OsO 4 scrubbing array served well in preventing OsO 4 from being released into the environment, until July of 2020 when a minor amount of 191 Os was found to have been released into the environment which resulted in a standby of building 4501 from July 2020 to February 2021.Since October 20 of 2020 a team of researchers from groups of RSTD and NEFD were organized to work for the following tasks: (1) Fully understand the root causes of the July incident of 191 Os release and create a redesigned 191 Os scrubbing array to eliminate the possibility of 191 Os release in future 188 W processes; (2) In Phase-1 tests, select a correct air (O 2 ) flow rate in 188 W process, that provides sufficient oxygen for the oxidation of tungsten metal, while still allowing sufficient resonance time for 191 Os removal by the scrubbers; (3) In Phase-2 tests, confirm the efficiency of the redesigned scrubbing array to absorb excessive amount of non-rad OsO 4 (generated from Os metal powder) at the selected air flow; (4) In Phase-3 tests, simulate the hot cell 188 W process by heating pressed non-rad W-Os pellets in a quartz vessel of new design and the new array under operation parameters selected in Phase-1 and -2; (5) Complete a new operation procedure for 188 W process in hot cell and a TM report as a summary of the Os Mitigation Project. This TM report summarizes the investigation on root causes from technical aspects in July 191 Os release and the correspondent improvement towards the redesigned 191 Os scrubbing system. The report presents the efforts made on selection of optimal operational parameter with the new scrubbing array and the quartz vessel of new design, based on experiment data obtained in Phase-1, -2 and -3 tests. Further renovations for 188 W process and additional improvement of W target treatment will also be discussed.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

Iodosodalite synthesis with hot isostatic pressing of precursors produced from aqueous and hydrothermal processes

Iodosodalite powders were synthesized using aqueous and hydrothermal methods, with or without sodium borosilicate glass binders, and then hot isostatically pressed (HIP) at 900 ? and 175 MPa for 3 h to convert into waste forms. After HIPing, the structures, compositions, morphologies, porosities, and leach rates of the aqueous and hydrothermally produced iodosodalite samples were compared. X-ray diffraction patterns of HIPed samples showed that iodosodalite remained as the dominant phase. However, the fraction of the iodosodalite phase decreased after HIPing, indicating decomposition of the iodosodalite structure during the HIP process. Scanning electron microscopy and chemical imaging on the cross sections of HIPed samples showed homogeneous elemental distribution for aqueous-grown iodosodalite, whereas hydrothermally grown iodosodalite samples had more heterogeneity and porosity. The densities of HIPed samples with more glass binder were generally higher than sampled HIPed without binder. The leach tests on samples containing 20 mass% glass binder showed that the iodine dissolution rate of HIPed hydrothermally grown iodosodalite was about 4 times higher than iodosodalite by aqueous method. This study provides alternative techniques for immobilizing iodine-streams in iodosodalite waste forms.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗