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Snyder, Michelle M. V.

Publications and source records attributed to Snyder, Michelle M. V..

5 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

200-DV-1 Laboratory Treatability Study: Proof-of-Principle Results

This document presents the Phase 1 technical approach and results of laboratory-scale treatability testing of nine in situ remedial technologies for their consideration in a future Feasibility Study (FS) for the 200-DV-1 OU. The primary objective of this initial assessment was a proof-of-principle testing evaluation primarily via batch experiments to determine the potential reduction and sequestration of primary contaminants of interest with and without potential co-contaminants of interest. An effectiveness rate of 35% transformation to immobile or nontoxic end products, along with other experimental indicators, was used by the project team to determine the technologies that advanced to Phase 2 for further evaluations. For most technologies, the performance was evaluated based on a series of sequential extractions designed to evaluate the mobility of contaminants before and after treatment, with each subsequent extraction representing a relative decrease in mobility. The results of this study will be used to inform testing for Phase 2 of the treatability study for further evaluation of selected technologies. The final results from the treatability study, following the completion of remaining experimental phases, will be used to determine whether the technologies tested can be appropriately evaluated in a FS to expand on the limited number of viable DVZ remediation technologies. After completion of the laboratory treatability study and the 200-DV-1 OU Remedial Investigation (RI) and Resource Conservation and Recovery Act Facility Investigation (RFI) of the waste sites, results will be evaluated to determine whether field studies are needed to provide additional information on effectiveness, implementability, or costs for evaluating these technologies in the FS for their site-specific application.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Stirred-Reactor Coupon Analysis: An International Round Robin Study

The objective of this task was to determine the precision of the SRCA technique when used to determine the dilute condition corrosion rate. To this end, an interlaboratory round robin study was conducted per the instructions in ASTM Practice E691 to measure the precision with which the SRCA test method can be conducted. Twelve independent labs from eleven different institutions each evaluated four glass compositions in three different conditions. The ASTM procedures recommend at least 6 labs participate in a round robin testing the same 3 materials in the same conditions to determine precision. In this case, 12 labs each performed 12 independent tests. This was only possible thanks to the multi-glass testing capability of the SRCA test. A total of 108 duplicate pairs were used to calculate the repeatability of the tests, with the same glass tested in the same vessel, producing as identical conditions as possible for the replicates. These test results were quite tightly clustered, with a median difference from the average value of the pair of only 2.9%. Based on the calculations outlined in ASTM E177-20 and a measured standard deviation of 4.74%, the intralaboratory repeatability limit (r) was calculated to be within 13.3% of the expected value with 95% confidence level. The reproducibility limit (R) of the test was examined using all 277 data points from the round robin. Because of the differences in dissolution rates due to pH variability and intrinsically for the 12 conditions tested, the reproducibility limits for each condition and overall were calculated from the percent relative residual value for each test. The SRCA test is expected to be reproducible within 64% of the expected value with a 95% confidence level.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Solubility controls on plutonium and americium release in subsurface environments exposed to acidic processing wastes

To identify the role of waste composition and sediment interactions in controlling Pu and Am mobility in contaminated sediments at Hanford, a legacy nuclear site, Pu and Am concentrations in solutions equilibrated with contaminated sediments from beneath the 216-Z-9 (Z-9) Trench were compared to the solubilities of PuO 2 materials, synthesized by methods representative of the disposed wastes, in the absence of sediments. Furthermore, this work shows that the solubilities of PuO 2 materials synthesized by different methods, and with varying particle sizes, agree with PuO 2 (am,hyd), although dissolution kinetics differed between materials. According to saturation index (SI) calculations, PuO 2 (am,hyd) is likely also controlling Pu release from sediments under conditions where phosphate concentrations are low. However, both Pu-phosphate and Am-phosphate phases, identified in SI calculations and by high resolution transmission electron microscopy, play roles in controlling release in low pH, high phosphate, shallow sediments just below the Z-9 Trench. The elevated phosphate is likely due to decomposition of tributyl phosphate from waste solutions over time. Sediments from deeper in the subsurface beneath the Z-9 Trench are less acidic and contain less phosphate, with Pu solubility likely controlled by PuO 2 (am,hyd) that precipitated following neutralization of the acidic waste stream. Controls on Am concentrations in deeper sediments are more complex and potentially involve sediment adsorption and/or release from Pu (1-x) Am x O 2 following Am in-growth. The concentrations of both Pu and Am were elevated in the colloidal fraction associated with shallow sediments, but not in PuO 2 experiments, suggesting the presence of Pu/Am pseudocolloids (e.g., Pu/Am associated with mineral colloids). However, Pu and Am association with the colloidal size fraction was not observed in deeper sediments, suggesting transport of Pu and Am to these depths beneath the Z-9 Trench was not due to colloidal transport.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Immobilizing Pertechnetate in Ettringite via Sulfate Substitution

Until recently, a nuclear waste form did not exist for the immobilization and long-term storage of the anionic form of radionuclide technetium-99 (99Tc), pertechnetate (TcO 4 - ). Across international nuclear waste sites, the inability to capture TcO 4 - poses a threat to the local environment due to the anion’s long half-life and environmental mobility under environmentally oxic conditions. To combat this challenge, most waste forms would capture 99Tc using materials that reduce TcO 4 - to Tc(IV) species, e.g. 99Tc oxides or 99Tc sulfides. However, this is a short-lived solution unless reducing capacities can be regenerated to last over the long life expectancy of some nuclear waste forms (~10,000 years). In light of recent experimental observations gained from the development of cementitious waste forms and unconfirmed hypotheses from the literature, this work explores how ettringite, a common mineral formed in cementitious materials with demonstrated success in incorporating oxyanions, may be used to immobilize TcO 4 - directly and over the life-time of the waste form. Using batch precipitation experiments, solid phase characterization techniques, and ab initio molecular dynamics simulations, we demonstrate herein successful incorporation of TcO 4 - into ettringite via substitution for SO 4 2- within the ettringite crystal structure. Here, these results are first of their kind and confirm the hypothesis that ettringite acts as an immobilizing mineral within cementitious waste forms for the highly mobile TcO 4 - oxyanion, enhancing the overall performance of these waste forms over time.

36 MATERIALS SCIENCE↗