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594 records · Page 33
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Weathering of chlorite from grain to watershed: The role and distribution of oxidation reactions in the subsurface

The reaction mechanism of weathering of chlorite, an important rock-forming phyllosilicate, is not well understood in natural settings. Here, in this work we investigated the weathering of Fe-rich chlorite from deep protolith to saprock to soil across a small shale-underlain watershed in the Appalachian Mountains, USA (Shale Hills). We found that oxidation of Fe(II) in chlorite always occurs prior to dissolution of the interlayers of the mineral. The oxidation of pyrite and chlorite commence near the water table across narrow depth intervals under the upper-catchment ridges, but well below the water table across wide depth intervals under the valley. We hypothesize that these patterns can be explained by hydrological and geochemical differences between the ridge and the valley: oxygenated water descends sub-vertically (1D flow) under the ridge, while under the valley, oxygen-depleted water moves upward to the stream and laterally out of the watershed in the subsurface (3D flow). Geochemical and mineralogical characterization indicates that the transformation of Fe-rich chlorite at Shale Hills is initiated by the oxidation of Fe(II). Next, the interlayer hydroxide sheet dissolves to form hydroxy-interlayered vermiculite and then vermiculite. During the transformation, Mg and Fe are released into solution and Fe is reprecipitated as goethite in pore space. Delivery of oxygen to the deep subsurface by infiltration of meteoric water is thought to control the initial transformation of chlorite at Shale Hills. It is possible that weathering of many Fe(II)-rich minerals is initiated by oxidation as mediated by rates of subsurface oxygen delivery.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Evolution of micro-pores in Ni–Cr alloys via molten salt dealloying

Porous materials with high specific surface area, high porosity, and high electrical conductivity are promising materials for functional applications, including catalysis, sensing, and energy storage. Molten salt dealloying was recently demonstrated in microwires as an alternative method to fabricate porous structures. The method takes advantage of the selective dissolution process introduced by impurities often observed in molten salt corrosion. This work further investigates molten salt dealloying in bulk Ni–20Cr alloy in both KCl–MgCl 2 and KCl–NaCl salts at 700 °C, using scanning electron microscopy, energy dispersive spectroscopy, and X-ray diffraction (XRD), as well as synchrotron X-ray nano-tomography. Micro-sized pores with irregular shapes and sizes ranging from sub-micron to several microns and ligaments formed during the process, while the molten salt dealloying was found to progress several microns into the bulk materials within 1–16 h, a relatively short reaction time, enhancing the practicality of using the method for synthesis. The ligament size increased from ~ 0.7 μm to ~ 1.3 μm in KCl–MgCl 2 from 1 to 16 h due to coarsening, while remaining ~ 0.4 μm in KCl–NaCl during 16 h of exposure. The XRD analysis shows that the corrosion occurred primarily near the surface of the bulk sample, and Cr 2 O 3 was identified as a corrosion product when the reaction was conducted in an air environment (controlled amount sealed in capillaries); thus surface oxides are likely to slow the morphological coarsening rate by hindering the surface diffusion in the dealloyed structure. 3D-connected pores and grain boundary corrosion were visualized by synchrotron X-ray nano-tomography. This study provides insights into the morphological and chemical evolution of molten salt dealloying in bulk materials, with a connection to molten salt corrosion concerns in the design of next-generation nuclear and solar energy power plants.

36 MATERIALS SCIENCE↗
DOE OSTI2020

TCCR Operational Summary and Optimization for Tank 9 Processing - 20314

Savannah River Remediation (SRR) manages and operates the liquid waste facilities at Savannah River Site (SRS) for the Department of Energy (DOE). Stored liquid waste is a complex mixture of insoluble solids (sludge) and soluble salts in an alkaline solution. SRR has deployed the Tank Closure Cesium Removal (TCCR) system, a tank-side ion exchange process, to remove radioactive cesium from salt waste and enable onsite disposal of the resulting decontaminated salt solution as low-level waste at the Saltstone facilities. The TCCR system consists of two prefilters, four ion exchange (IX) columns, one resin trap, and a ventilation system. The IX process uses a form of inorganic crystalline silicotitanate (CST), which has a high affinity for cesium and other alkali metals, strontium, and actinides. This process is currently deployed utilizing salt feed from Tank 10, with future plans to dissolve solid salt in Tank 9 and transfer the salt solution to Tank 10 for processing through TCCR. The feed for TCCR must be created from salt-cake in Tank 10 through a dissolution process. Once enough salt has been dissolved, a qualification process is entered. This process characterizes the feed and ensures the cesium loading on the columns will not cause boiling of waste within the columns during or after processing. Once the batch has been qualified, salt waste is fed to the TCCR system through a transfer pump in the center of the tank. The waste is filtered through a set of two shielded, dead-end prefilters that prevent solids buildup in the columns. The filtered salt solution then travels to the shielded IX columns, which can be operated individually or in series, where the cesium is sorbed on the CST media. The decontaminated salt solution (DSS) then travels through a resin trap and out of the module to Tank 11. TCCR has successfully processed approximately 795,000 L of Tank 10H radioactive salt waste over two batches to date. There has not yet been a system induced shutdown. The prefilters performed as expected with only minor degradation in recovery of differential pressure after a backflush sequence. The time between backflushes decreased as each batch reached the end of processing. The hydraulics in the IXCs mostly performed as expected at all flow rates, except for one IXC that will be further investigated during Batch 3 processing. The TCCR system has shown some opportunities for more efficient processing during the length of the demonstration so far. For future processing of material from Tank 9H through Tank 10H and the TCCR unit, TCCR 1A will implement changes to the prefilters and the IXCs. The prefilters will have an increased surface area and a new filter media in an effort to increase time between filter swaps and improve backwashing cleaning capability. The IXCs will have a reduced diameter to allow for increased heat transfer out of the column and increased loading of Cs-137. Additionally, a new form of CST with an increased kinetic performance is being investigated for use during TCCR 1A operation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2014

Probabilistic Assessment of Hypobaric Decompression Sickness Treatment Success

The Hypobaric Decompression Sickness (DCS) Treatment Model links a decrease in computed bubble volume from increased pressure (DeltaP), increased oxygen (O2) partial pressure, and passage of time during treatment to the probability of symptom resolution [P(symptom resolution)]. The decrease in offending volume is realized in 2 stages: a) during compression via Boyle's Law and b) during subsequent dissolution of the gas phase via the O2 window. We established an empirical model for the P(symptom resolution) while accounting for multiple symptoms within subjects. The data consisted of 154 cases of hypobaric DCS symptoms along with ancillary information from tests on 56 men and 18 women. Our best estimated model is P(symptom resolution) = 1 / (1+exp(-(ln(Delta P) - 1.510 + 0.795×AMB - 0.00308×Ts) / 0.478)), where (DeltaP) is pressure difference (psid), AMB = 1 if ambulation took place during part of the altitude exposure, otherwise AMB = 0; and where Ts is the elapsed time in mins from start of the altitude exposure to recognition of a DCS symptom. To apply this model in future scenarios, values of DeltaP as inputs to the model would be calculated from the Tissue Bubble Dynamics Model based on the effective treatment pressure: (DeltaP) = P2 - P1 | = P1×V1/V2 - P1, where V1 is the computed volume of a spherical bubble in a unit volume of tissue at low pressure P1 and V2 is computed volume after a change to a higher pressure P2. If 100% ground level O2 (GLO) was breathed in place of air, then V2 continues to decrease through time at P2 at a faster rate. This calculated value of (DeltaP then represents the effective treatment pressure at any point in time. Simulation of a "pain-only" symptom at 203 min into an ambulatory extravehicular activity (EVA) at 4.3 psia on Mars resulted in a P(symptom resolution) of 0.49 (0.36 to 0.62 95% confidence intervals) on immediate return to 8.2 psia in the Multi-Mission Space Exploration Vehicle. The P(symptom resolution) increased to near certainty (0.99) after 2 hrs of GLO at 8.2 psia or with less certainty on immediate pressurization to 14.7 psia [0.90 (0.83 - 0.95)]. Given the low probability of DCS during EVA and the prompt treatment of a symptom with guidance from the model, it is likely that the symptom and gas phase will resolve with minimum resources and minimal impact on astronaut health, safety, and productivity.

Conkin, Johnny↗
DOE OSTI2019

The Microstructure, and Mechanical and Corrosion Properties of As-Cast and As-Extruded Mg-2%Zn-x%Cu Alloys After Solution and Aging Heat Treatments

The effects of different Cu concentrations and heat treatment on the microstructure, mechanical and corrosion properties of the as-cast and hot-extruded Mg-2%Zn alloys were studied. The results showed that T6 heat treatment led to the dissolution of intermetallics and breaking the continuous networks of intermetallics. Also, the grain size of the base alloy decreased with further Cu additions after solution and aging heat treatments. Microstructural characterization indicated that Mg(Zn,Cu) and Mg(Zn,Cu){sub 2} intermetallics were formed which their volume fractions increased with exceeding Cu additions. Moreover, the addition of 3 wt.% Cu and T6 heat treatment increased the Brinell hardness of both as-cast and hot-extruded specimens reached to 94 HBN and 116 HBN, respectively. The optimum amounts of the ultimate tensile strength and elongation values obtained from 0.1 wt.% Cu addition for both conditions after T6 treatment which were 253 MPa, 11.5% and 263 MPa, 14.7%, respectively. However, more Cu addition (> 0.5 wt.%) deteriorated the tensile properties of the alloy due to the high volume fraction of intermetallics. The yield tensile strength increased continuously with different Cu additions for the cast and extruded alloys after solution and aging heat treatments. After T6 heat treatment, extruded alloys showed better anti-corrosion properties than those of the cast alloys, while the extruded Mg-2%Zn-0.1%Cu alloy exhibited the best anti-corrosion property. Further Cu addition increased the volume fraction of intermetallics and enhanced the corrosion rate which was due to the galvanic couple effect.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Characterization of Contaminants and Mobility in WMA A-AX Boreholes D0006 and D0008

This study was initiated to determine if subsurface contaminants in the A-AX vadose zone at the Hanford site (boreholes D0006 and D0008) caused corrosion of nearby stainless steel well casings (299-E-24-19, 299-E25-46, and 299-E25-236) and, based on the type and distribution of contaminants, determine if contaminants are from one or more sources including a) the A-104 and/or A-105 tanks, b) the 242A evaporator, and c) the 200-E-286 ditch. A tiered approach was used for characterizing contaminants in vadose zone sediments starting with mobile and total contaminant concentrations (Tier 1), characterizing contaminant speciation (Tier 2), and measuring contaminant leaching (Tier 3). Total and mobile contaminant measurements at seven depths showed elevated sulfate and nitrate in all samples (130 to 280 ft depth), elevated chloride in five depths, and no elevated radionuclides except I-129 in D0006 at 206 to 208 ft depth. If these contaminants were initially acidic, stainless-steel corrosion would occur. However, because sediments likely have a high acid neutralization capacity, an acidic spill would have been neutralized within tens of feet, so casing corrosion would likely occur in shallow sediments. GEL labs data also shows low (but above natural) chromate from shallow to 170 ft depth, then higher chromate to 280 ft, suggesting a surface Cr spill or casing corrosion. GEL labs data also showed evidence of more than one plume, with deep nitrate, sulfate, and chloride migration, but only shallow tritium migration. Tritium should migrate nearly unretarded in the subsurface like nitrate and chloride. Elevated aqueous cations and anions in the D0006 206 to 208' depth pore water (SO 4 2- , NO 3 - , Na + , K + , Ca 2+ , Mg 2+ , Si) could be the result of acid disposal and calcite dissolution at shallow depth, with deeper migration of SO 4 2- , NO 3 - from acids and Ca 2+ and Mg 2+ from calcite. Elevated Si (160 mg/L) indicates dissolution of clays or other silicates, which can also occur in acids. Acid extractable metals did show elevated Cr (but not Fe, Ni, Mo, and Mn) above natural levels, which may indicate stainless-steel well corrosion or transport of metals from a shallow spill. The elevated I-129 in D0006 at 206 to 208' depth was low with 0.34 μg/g in pore water and 1.5 μg/g total extractable I-129 in the sediment, which indicates significant fraction of I-129 was bound in one or more precipitates. Although total I-129 was below detection limits in most extractions due to high molybdate interference, I-127 analysis provided some insight into potential precipitate phases. Total I-127 was mainly extracted by acetate or acetic acid (67%), which may be iodine in calcite. Inorganic carbon measurement identified 1.65% calcite, and total organic carbon (TOC) was below detection limits, so iodine was not associated with organic matter. Nearly all the I-127 was present as iodide. In 1-D leach experiments, I-127 as iodide leached from the sediment within 2 pore volumes. Due to high and dynamic molybdate leaching from the sediment and interference with I-129 analysis, there were few quantifiable I-129 measurements. The calculated molybdate release rates from the sediment suggested Mo was being released from the same precipitate phase, such as molybdate incorporated into calcite. Overall, because there was little radioactive contamination measured in D0006 and D0008 boreholes, potential releases from A-104 and A-105 did not reach this location south of the A tank farm. Elevated cations and anions found in D0006 and D0008 at depth as the result of acidic disposal could cause shallow stainless steel casing corrosion. The nearby 242A evaporator generally contained alkaline waste and tritium, but the nearby 200-E-286 ditch generally contained high Cl - liquid effluent, so these are unlikely to be acidic sources.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Dual in-aquifer and near surface processes drive arsenic mobilization in Cambodian groundwaters

Millions of people globally, and particularly in South and Southeast Asia, face chronic exposure to arsenic from reducing groundwaters in which. Arsenic release to is widely attributed largely to reductive dissolution of arsenic-bearing iron minerals, driven by metal reducing bacteria using bioavailable organic matter as an electron donor. However, the nature of the organic matter implicated in arsenic mobilization, and the location within the subsurface where these processes occur, remains debated. In a high resolution study of a largely pristine, shallow aquifer in Kandal Province, Cambodia, we have used a complementary suite of geochemical tracers (including 14 C, 3 H, 3 He, 4 He, Ne, δ 18 O, δD, CFCs and SF 6 ) to study the evolution in arsenic-prone shallow reducing groundwaters along dominant flow paths. The observation of widespread apparent 3 H- 3 He ages of <55years fundamentally challenges some previous models which concluded that groundwater residence times were on the order of hundreds of years. Surface-derived organic matter is transported to depths of >30m, and the relationships between age-related tracers and arsenic suggest that this surface-derived organic matter is likely to contribute to in-aquifer arsenic mobilization. A strong relationship between 3 H- 3 He age and depth suggests the dominance of a vertical hydrological control with an overall vertical flow velocity of ~0.4±0.1m·yr-1 across the field area. A calculated overall groundwater arsenic accumulation rate of ~0.08±0.03μM·yr -1 is broadly comparable to previous estimates from other researchers for similar reducing aquifers in Bangladesh. Although apparent arsenic groundwater accumulation rates varied significantly with site (e.g. between sand versus clay dominated sequences), rates are generally highest near the surface, perhaps reflecting the proximity to the redox cline and/or depth-dependent characteristics of the OM pool, and confounded by localized processes such as continued in-aquifer mobilization, sorption/desorption, and methanogenesis.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2019

Root-driven weathering impacts on mineral-organic associations in deep soils over pedogenic time scales

Plant roots are primary weathering agents in deep soils, yet the impact of resulting mineral transformations on the vast deep soil carbon (C) reservoir are largely unknown. Root-driven weathering of primary minerals causes the formation of reactive secondary minerals, which may protect mineral-organic associations (MOAs) for centuries or millennia. Conversely, root-driven weathering may also transform secondary minerals, potentially enhancing the bioavailability of C previously protected in MOAs. Here we examined the impact of root-driven weathering on MOAs and their capacity to store C over pedogenic time scales. To accomplish this, we compared soil that experienced root-driven weathering, resulting in the formation of discrete rhizosphere zones in deep soil horizons (100-160 cm) of the Santa Cruz Marine Terrace chronosequence (65ka-226ka), with adjacent soil that experienced no root growth. Using a combination of radiocarbon, mass spectrometry, Mössbauer spectroscopy, and X-ray spectromicroscopy approaches, we characterized transformations of mineral-organic associations in relation to changes in C content, turnover and chemistry across four soils ranging in age from 65 to 226 ka. We found that the onset of root-driven weathering (65-90ka) increased the amount of C associated with poorly crystalline iron (Fe) and aluminum (Al) phases, particularly highly-disordered np-goethite. This increase coincided with greater C concentrations, longer C residence times, and a greater abundance of microbially-derived C. Continued root-driven weathering (137-226ka) did not significantly change the amount of C associated with crystalline Fe and Al phases, but resulted in a decline in the amount of C associated with poorly crystalline metal phases. This decline coincided with a decrease in C concentrations and potential turnover rates, and a shift toward plant-derived C. In contrast, soil not affected by root-driven weathering showed comparatively low amounts of C bound in poorly crystalline Fe and Al phases regardless of soil age and, correspondingly, lower C concentrations and estimated residence times. Furthermore, our results demonstrate that root driven-formation and disruption of poorly crystalline Fe and Al phases is a direct control on both C accrual and loss in deep soil. This finding suggests that root impacts on soil C storage are dependent on soil weathering stage, a consideration that is critical for future predictions of the vulnerability of deep soil C to global change.

58 GEOSCIENCES↗
DOE OSTI2018

Integrated characterization of CO 2 storage reservoirs on the Rock Springs Uplift combining geomechanics, geochemistry, and flow modeling

This project improves understanding of the effects CO 2 injection and storage on geomechanical, petrophysical, and other properties on two potential reservoirs, the Madison Limestone and Weber Sandstone, at the Rock Springs Uplift, southwest Wyoming. Rock physics studies, laboratory experiments, and computer simulations were employed. Two groups of rock samples were prepared, one aged with formation brine and the second aged with CO 2 -saturated formation brine and subsequently used for geomechanical, geochemical, and petrophysical testing. Rock physics studies provided a facies classification to discriminate among different lithologies and to investigate the feasibility of predicting petrophysical properties using seismic inversion. By applying a zone-wise regression relation established in this facies analysis, a 3D reservoir model was developed. It was determined that time-lapse seismic data are not feasible for monitoring CO 2 displacement at the Rock Springs Uplift or to mitigate leakage-risks situations because pressure and fluid saturation have small effects on elastic velocities and synthetic seismograms. The small variations that do occur are due to the low porosity of the sandstone and the high stiffness of the dolomite. In the Madison Limestone, mineral dissolution due to reaction with CO 2 -saturated brine was observed in highly permeable pathways while mineral precipitation happened in intergranular or micropores. Changes in pore size were quantified via Time-Domain Nuclear Magnetic Resonance transverse relaxation time (T 2 ) and diffusion coefficient distributions. As opposed to continuous injection of CO 2 -saturated brine, a subtle increase in porosity after static ‘soaking’ with CO 2 -saturated brine increases permeability due to the heterogeneity of changes in pore and pore throat sizes. Porosity-permeability relationships are fitted using a power-law function that can be adopted to model time-evolved porosity and permeability in a transport-limited CO 2 injection process. The effect of CO 2 -saturated brine on the linear elastic properties of Weber Sandstone and Madison Limestone cannot be generally concluded. Considering horizontal and vertical orientations and different confining pressures, no consistent relationship between stress (or strain) data and reaction with CO 2 -saturated brine was observed in the nonlinear plastic and post-failure regime. The change of elastic constants due to reaction with CO 2 -saturated brine is more significant in Madison Limestone than in Weber Sandstone; however, no consistent trend was observed. Reservoir simulations suggest that injectivity for Weber Sandstone is adversely impacted by low permeability. Changes in confining pressure do not impact injection rate and can be ignored while the effect of changes to pore pressure cannot. The effect of confining pressure and pore pressure on the size of the CO 2 plume in the Madison Limestone is negligible because of higher permeability. Geomechanical and petrophysical properties induced by CO 2 -brine-rock reactions were time-dependent. Simulations considered these time-dependent properties and indicate that injectivity, porosity distribution, pressure distribution and CO 2 saturation were affected.

58 GEOSCIENCES↗
DOE OSTI2020

Investigation of Thermolytic Hydrogen Generation Rate in Tank 44 Dissolved Saltcake Samples

Saltcake core samples collected from Tank 44 in 2006 were dissolved to provide material for HGR measurements applicable to F-Area dissolved saltcake material. Additionally, characterization was performed on the Tank 44 saltcake material. The following are key results from the Tank 44 saltcake characterization. The Tank 44 Upper Saltcake Composite, corresponding to the 171 to 285 inch tank level, contained by mass approximately 69% sodium nitrate, 11% sodium carbonate, 8% sodium nitrite, smaller amounts of other salts and components, and 9% unquantified (which includes water, water of hydration, oxygen/hydrogen content of oxides and hydroxides, and uncertainty). The Tank 44 Lower Saltcake Composite, corresponding to the 76 to 114 inch tank level, contained by mass approximately 49% sodium carbonate, 18% sodium nitrate, smaller amounts of other salts, at least 8% sludge, and 9% unquantified (see above). The dissolved saltcake contained free hydroxide less than quantifiable (<0.01 M) due to the limited quantity of material that could be removed from the Shielded Cells based on the sample radioactivity. Measurement by pH paper provided an approximate pH of 12. The following are key results from the Tank 44 HGR testing. During boiling at 106.7 °C, HGR for Tank 44 dissolved saltcake without added glycolate was 7.2×10 -8 ft 3 h -1 ga l-1 . During boiling at 106.9 °C, HGR for Tank 44 dissolved saltcake with 1000 mg/L of added glycolate was 8.2×10 -8 ft 3 h -1 gal -1 . ∙ For the test without added glycolate, the first several HGR measurements at 70, 85, and 100 °C gave indication of the release of dissolved hydrogen and should not be used to represent the sustained thermolytic HGR for those temperatures. The measurements at boiling are the best representation of thermolysis in this testing. Carbon dioxide was observed at concentrations up to 6 vol% in the flow-system offgas for the test at boiling. ∙ Methane generation was observed at 100 °C and boiling. Methane concentration in the total gas generated during testing remained well below the lower flammability limit for methane in air. The addition of 1000 mg/L of glycolate did not have a significant impact on the hydrogen generation rates measured during this testing. The low hydroxide concentration in the Tank 44 dissolved saltcake likely influenced the relatively low thermolytic HGR and high carbon dioxide release observations in this testing. Based on the observation that methane was generated or released upon heating SRS radioactive Tank 44 waste samples to 100 °C and above, we recommend gaining a greater understanding of the cause and mechanism of its generation. First, the applicable literature should be reviewed to reveal the thermolytic methane generation mechanisms of possible methane generating species in the SRS CSTF. If warranted, a plan should be developed for simulant tests with methylated siloxanes and other applicable compounds in order to gain a better mechanistic understanding of methane generation in the SRS CSTF.

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

Investigation of Thermolytic Hydrogen Generation Rate in Tank 44 Dissolved Saltcake Samples

Saltcake core samples collected from Tank 44 in 2006 were dissolved to provide material for HGR measurements applicable to F-Area dissolved saltcake material. Additionally, characterization was performed on the Tank 44 saltcake material. The following are key results from the Tank 44 saltcake characterization. The Tank 44 Upper Saltcake Composite, corresponding to the 171 to 285 inch tank level, contained by mass approximately 69% sodium nitrate, 11% sodium carbonate, 8% sodium nitrite, smaller amounts of other salts and components, and 9% unquantified (which includes water, water of hydration, oxygen/hydrogen content of oxides and hydroxides, and uncertainty). The Tank 44 Lower Saltcake Composite, corresponding to the 76 to 114 inch tank level, contained by mass approximately 49% sodium carbonate, 18% sodium nitrate, smaller amounts of other salts, at least 8% sludge, and 9% unquantified (see above). The dissolved saltcake contained free hydroxide less than quantifiable (<0.01 M) due to the limited quantity of material that could be removed from the Shielded Cells based on the sample radioactivity. Measurement by pH paper provided an approximate pH of 12. The following are key results from the Tank 44 HGR testing. During boiling at 106.7 °C, HGR for Tank 44 dissolved saltcake without added glycolate was 7.2×10 -8 ft 3 h -1 gal -1 . During boiling at 106.9 °C, HGR for Tank 44 dissolved saltcake with 1000 mg/L of added glycolate was 8.2×10 -8 ft 3 h -1 gal -1 . For the test without added glycolate, the first several HGR measurements at 70, 85, and 100 °C gave indication of the release of dissolved hydrogen and should not be used to represent the sustained thermolytic HGR for those temperatures. The measurements at boiling are the best representation of thermolysis in this testing. Carbon dioxide was observed at concentrations up to 6 vol% in the flow-system offgas for the test at boiling. Methane generation was observed at 100 °C and boiling. Methane concentration in the total gas generated during testing remained well below the lower flammability limit for methane in air. The addition of 1000 mg/L of glycolate did not have a significant impact on the hydrogen generation rates measured during this testing. The low hydroxide concentration in the Tank 44 dissolved saltcake likely influenced the relatively low thermolytic HGR and high carbon dioxide release observations in this testing. Based on the observation that methane was generated or released upon heating SRS radioactive Tank 44 waste samples to 100 °C and above, we recommend gaining a greater understanding of the cause and mechanism of its generation. First, the applicable literature should be reviewed to reveal the thermolytic methane generation mechanisms of possible methane generating species in the SRS CSTF. If warranted, a plan should be developed for simulant tests with methylated siloxanes and other applicable compounds in order to gain a better mechanistic understanding of methane generation in the SRS CSTF.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Nanoscale observations of Fe(ii)-induced ferrihydrite transformation

Because of its sorption properties, transformation of the nanomineral ferrihydrite (Fh) into more stable lepidocrocite (Lp) or goethite (Gt) has important impacts on the fate of metals, nutrients, and contaminants in soils/sediments. Although it is well known that the transformation rate is greatly accelerated under suboxic conditions by aqueous Fe(II), the enabling mass transfer process remains an ongoing debate among various mechanisms including dissolution/reprecipitation, solid-state recrystallization, and particle-mediated growth. In this paper, using electron microscopy, we examine the mineralogical evolution of 2-line Fh to Lp/Gt catalyzed by Fe(II) under strict anoxic conditions, including evaluation of Cl–SO 4 –HCO 3 anion effects. Emergence of Lp/Gt crystallites at the nanoscale was observed at ~20 min of reaction, earlier than previously reported. Lp is the first phase to nucleate in Cl-rich solutions without HCO 3 – ; whereas Lp and Gt concomitantly nucleate in SO 4 2– -rich solutions, and also when co-solute HCO 3 – is added. Lp crystallites nucleate as quasi-2D nanosheets one-unit-cell thick that contour the Fh surface; in contrast, rod-shaped (in Cl/SO 4 ) or acicular needle-shaped (in HCO 3 ) Gt crystals nucleate and grow radially outward from the Fh aggregates. Stages of transformation monitored by in situ μ-XRD coupled with aqueous Fe(II) uptake/release measurements are correlated with a short initial sorption stage followed by the onset of Lp/Gt growth that then progresses to Lp loss in favor of Gt. Microscopy data overwhelmingly support dissolution/reprecipitation as the underlying mechanism, including direct evidence for classical ion-by-ion Lp/Gt growth and Lp dissolution. The collective findings imply that the iron mass transfer through solution to distal Lp/Gt growth fronts is a critical enabling process facilitating rapid transformation.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2023

Produced Fluid Induced Mineralogy and Elemental Alterations of Caney Shale, Southern Oklahoma

ABSTRACT This study involves batch reactor experiments and subsequent analyses of samples from Caney Shale in the Ardmore Basin of South-Central Oklahoma. Samples include mainly rock cores and cuttings recovered from two wells respectively drilled vertically through and horizontally across the Caney Shale. Mineralogical compositions are obtained by X-Ray Diffraction (XRD) measurements whilst microstructure and elemental distribution are acquired by Scanning Electron Microscopy/Energy Dispersive Spectroscopy (SEM/EDS) respectively. Batch experiments are then conducted using selected rock samples and produced fluid from the Caney Formation. Deionized water is also reacted with some samples to serve as standard. Experiments are conducted at 95°C and ambient pressure for 7 and 30 days to assess the geochemical rock-fluid interactions. Results show rock mineralogical compositions are predominantly quartz, feldspar, carbonates, and clay with minor pyrite. Post-experimental mineralogical changes observed in samples include increased amorphous entities especially within the clay portions of XRD plots and dissolution of feldspar and carbonate minerals and formation of new mineral phases, mostly clays and salts. These are corroborated by EDS elemental analyses which show decreased elemental compositions. The implications of reactions mentioned above include but not limited to, scale formation, clay fines migration and shale softening all of which pose significant permeability impairment on formation over time. INTRODUCTION Shale reservoirs account for a large share of unconventional reservoirs in the world (Lyu et al., 2015). However, ultra-low permeability and high clay compositions pose significant challenges when producing from these reservoirs (Dawuda and Srinivasan, 2022, 2023). Producing from these reservoirs therefore requires horizontal drilling and hydraulic fracturing technologies which have proven their efficacy in generating substantial permeability in reservoirs to ensure production (Fujian et al., 2019; Liu et al., 2018). Even after expensive horizontal drilling and hydraulic fracturing, geochemical reactions between engineered fluids and formation leads to fracture constriction and adversely impact petrophysical properties (permeability and porosity) of the reservoir. These technologies are therefore under constant development and improvement in various aspects to ensure fine tuning for specific reservoirs. Under present conditions, much of the hydrocarbon reserves in unconventional shale reservoirs are left unproduced due to rapid decline in permeability following resumption to production after hydraulic fracturing. It is therefore essential to understand the range of geochemical reactions that cause rapid depletion of permeability after hydraulic fracturing and apply these to each shale reservoir to ensure substantial recovery rates.

Awejori, G. A.↗
DOE OSTI2019

Dissolution of Light Curium Oxide with a Catalyzed Electrolytic Process

The ability to produce transcurium elements and isotopes, such as 249 Bk and 252 Cf, requires the appropriate feed material and a high thermal neutron flux. Currently, at Oak Ridge National Laboratory (ORNL), the High Flux Isotope Reactor (HFIR) provides one of the world’s highest steady state thermal neutron fluxes. The feed material consists of recycled curium from the decades of production of 249 Bk, 252 Cf, 253 Es, and 255/257 Fm at the Radiochemical Engineering Development Center (REDC). Over time, the feed material has been depleted and needs to be replenished. The current feed material is called “heavy recycled curium” and consists of over 50 wt% of 246/248 Cm isotopes. In order to replenish this stock material, additional curium is needed. The only curium material that is currently available consists predominantly (greater than 80 wt%) of 244 Cm. This material also contains 240 Pu from the radioactive decay of the 244 Cm. Because of the amount of 244 Cm is greater than 50% of the total Cm, this material is termed “light curium.” Today, 249 Bk and 252 Cf are produced in only two places in the world: the United States and Russia. In the United States, the US Department of Energy and its predecessors have made these isotopes and others available to research agencies and commercial users since the late 1960s as part of the Department of Energy heavy-element production program located at ORNL. Currently these materials are provided through the DOE Office of Science Nuclear Physics-Isotope Development and Production for Research and Applications and the Basic Energy Science-Heavy Element Chemistry. Feedstock for the production of 252 Cf and other transcurium element by-product isotopes was originally prepared by means of irradiation of multi-kilograms of 239 Pu in the large production reactors at the Savannah River Site. Products of the Savannah River Site irradiation were predominantly 242 Pu, 243 Am, and 244 Cm. These materials were shipped to ORNL for production of multi-milligram amounts per year of 252 Cf and other transcurium element isotopes by means of irradiation in the HFIR, followed by separation and purification processing in the Radiochemical Engineering Development Center (REDC). The transcurium element isotopes produced were distributed to other US and collaborating foreign laboratories for research studies to determine the nuclear and chemical properties of these rare materials. The Cm feed material plays a key role in the efficient production of 252 Cf. The early targets irradiated in HFIR contained only the 242 Pu, 243 Am, and 244 Cm produced at Savannah River Site, and the rate of production of 252 Cf was less than 1 mg per 8 g of target material as shown in Figure 1. Also, shown in Figure 1 is the significant difference in 252 Cf production between the light and heavy curium; it differs by approximately 2 orders of magnitude.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Striking stability of a mixed-valence thallium(III)-thallium(I) complex in some solvents

At the dissolution of solid anhydrous thallium(III) trifluoromethanesulfonate, Tl(CF 3 SO 3 ) 3 , or thallium(III) trifluoroacetate, Tl(CF 3 COO) 3 , in dimethylsulfoxide (dmso) or N,N,N’,N’-tetramethylurea (tmu), intensely red-colored complexes are formed. This red thallium complex is stable for years in dmso, while it is reduced fairly rapidly to thallium(I) in tmu with a half-life time of an hour. At the dissolution of Tl(CF 3 SO 3 ) 3 in N,N-dimethylpropyleneurea (dmpu) an immediate reduction to thallium(I) takes place. A stable colorless aqueous thallium(III) solution is obtained at the dissolution in acidic water. Stable dmso solutions and solid dmso solvates of thallium(III) perchlorate, nitrate and trifluormethanesulfonate can be prepared by adding dmso to concentrated acidic aqueous thallium(III) solutions. These experimental observations conclude that the pure solids Tl(CF 3 SO 3 ) 3 and Tl(CF 3 COO) 3 play an essential role in the formation of the red-colored thallium complexes. 205 Tl NMR data show that the red thallium complex contains equal amounts of thallium(III) and thallium(I). The structure of the red thallium complex in dmso, as determined by EXAFS, has Tl–O bond distances of 2.216(3) and 2.80(2) Å, which are in very close agreement with the bond distances obtained in the pure dmso solvates of the thallium(III) and thallium(I) ions, respectively, and a Tl···Tl distance of 3.49(1) Å bridged by oxygen atoms. From the EXAFS data it is impossible to distinguish if dmso molecules and/or trifluoromethanesulfonate ions act as bridges. DFT calculations could eliminate some structures due to the irrelevant structural parameters or the energetics of the proposed reactions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Finding ways to reduce nuclear waste: searching for the unknown one step at a time

In my home country, Venezuela, research has been stagnant. Due to the political turmoil and the crisis, many educated people have left the country in search for a better life. This has caused a deficit in any technological and scientific advances, making Venezuela one of the first South American countries to have its rate of publications decline by 29% in 2013. Currently, Venezuela lacks the infrastructure and the means to keep up with the research progress as compared to other countries in South America, such as Brazil. Since coming to the United States (US), and currently working for a national laboratory, the active research environment endorses a wide range of careers and engineering programs that allow researchers to thrive at any given field. Researchers have access to funds and tools to succeed in developing materials for the future. There are 17 national laboratories in the US, and all of these have a different research focus/objective. As examples, Los Alamos National Laboratory and Sandia National Laboratory focus is on national homeland security, weapon science, radiation effects, among others. Argonne National Laboratory focuses on nuclear energy, energy storage, high performance computing, etc. At Idaho National Laboratory (INL) the research focuses on innovating nuclear energy and clean energy resources, critical infrastructure materials, along with fuel cycle solutions to manage, dispose and find ways to recycle current and future radiological waste. Compared to other national laboratories, INL focuses slightly more on applied processes and how nuclear energy can be innovated to next reactor design and technologies. The research being conducted at INL made me apply for a Seaborg distinguished postdoctoral position. For the position itself, the researcher must submit a proposal related to actinide chemistry on a research field area. In this position, 50% of my time will be focused on my own proposal. The proposal that I am working on is focused on the innovation of nuclear energy and fuel cycle recycling, which is why I was mainly interested on working at this national laboratory. To give a bit more context of what my proposal is about, a little bit of background is necessary: After the nuclear fuel (UO2) is used in a reactor, the fuel matrix is then characterized by various fission products (FP). Among these FP (including rare earth elements, alkali/alkaline earths, and actinides), many can potentially be recovered through nuclear reprocessing technologies. In pyroprocessing, the used nuclear fuel undergoes electrochemical dissolution into a molten chloride salt mixture in an electrorefiner. Initially, uranium is reduced onto an inert cathode by applied potentials. However, numerous remaining FPs accumulate in the melt and pose challenges for recovery by an inert electrode, particularly the rare earth elements (e.g., Nd, Gd, Pr, Sm) due to their multivalent oxidation states and tendencies toward side reactions, leading to their dissolution in the electrolyte. These recovery challenges result in inefficiencies and necessitate the continual discarding of the molten chloride salt, thereby generating additional waste. Furthermore, the presence of rare earth elements and other fission products in the molten salt electrolyte alters its physical and chemical properties, affecting both uranium recovery efficiency and the longevity of the molten chloride salt. To improve the recovery efficiency of the FP, specifically rare earth elements, I am investigating the fundamental interactions between rare earth elements in the molten chloride salt and their metallic form. The kinetic pathways and the chemical reactions of these elements will give insights on how the recovery efficiency can be improved. The interactions and speciation of these elements are being studied by spectro-electrochemistry at high temperature environments in quartz and other ceramic materials (e.g., alumina crucibles). Some of the challenges I am facing specifically relates the reactivity of some of these elements with different glass and crucible materials. Although my research focuses on fundamental science, it will benefit the applied process by generating new scientific knowledge and closing the gap for an efficient recycling of the waste: one step at a time.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Smoothed particle hydrodynamics modeling and analysis of oxide reduction process for uranium oxides

A common kinetic feature for oxide reduction chemical/electrochemical processes is oxygen transport via a porous metallic layer, which has been considered as a rate-determining step for reducing uranium oxides to metallic uranium. Accounting this kinetic behavior must involve the resolution of the moving reactive interface between shrinking oxide and expanding metal phases. This study presents a numerical model using smoothed particles hydrodynamics (SPH) to effectively deal with the evolution of the shrinking core reaction interface and oxygen transport via mass transfer of lithium oxide (Li 2 O) species in multiple mass transfer domains. We successfully validated the proposed model against a theoretical derivation for the oxide reduction process handling a shrinking oxide core with molten salt and metal ash medium on a simple planar geometry. Armed with successful validation results, the model examined a realistic reactant geometry to extend the arguments beyond the one-dimensional analyses, allowing the proposed model to apply to general application scenarios with multi-dimensional geometry. Here, this study demonstrated that the proposed model could simulate and evaluate an arbitrary reaction basket design without iterative experimental trials, which is prohibitive for a scaled high-temperature molten salt study in an inert environment. The construct potentially provides not only deep insights on multiphysics behaviors governing the process dynamics but also a robust framework for evaluating and screening candidate basket designs in the most cost-effective manner.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Technical Evaluation of Accelerated Basin De-Inventory Material Addition to Sludge Batch 11 (Rev. 1)

The Accelerated Basin De-inventory (ABD) program involves discarding spent nuclear fuel that is currently stored in L-Basin to the Defense Waste Processing Facility (DWPF) for vitrification. The first ABD discards will occur during the preparation of Sludge Batch (SB) 11. Savannah River Mission Completion has requested that the Savannah River National Laboratory assess the technical gaps related to the increased gadolinium poisoning requirement and the impacts of performing the Low Temperature Aluminum Dissolution (LTAD) process in Tank 51 with H-Canyon discards present. The following summarizes the evaluation of the impacts of increasing the quantity of gadolinium (and related topics) from what was previously evaluated in the SRNL studies of gadolinium-poisoned ABD material solubility, the overall ABD flowsheet review, and increasing the fissile mass loading in glass: 1) Based on literature surveys, there is no indication that organic interactions with gadolinium will be significant at the high pH (typically >13) conditions of the Concentration, Storage, and Transfer Facilities. Any interactions of gadolinium with organics in DWPF are not expected to adversely impact DWPF or downstream facilities. Thus, there is little-to-no residual risk from organic interactions with gadolinium [Gap closed]; 2) Adding depleted uranium to ABD material, targeting 235 U enrichment of 4.90% within each transfer window, will mitigate potential impacts from an increase in soluble 235 U enrichment during sludge washing and LTAD. The plan to take advantage of previous transfers and allow 235 U enrichment of >5% during the final transfer window carries a risk that Tank 51 supernate will have a 235 U enrichment of >5%, which should be evaluated for acceptance; 3) Increasing the gadolinium mass ratio to 3.0:1 Gd: 235 U(eq SLU ) should lead to the same or higher partitioning of gadolinium into the solid phase within the DWPF Chemical Process Cell, resulting in both liquid and solid phases with expected partitioning of Gd consistent with the prior solubility study [Gap closed for SB11]; 4) There are no expected impacts on DWPF melt temperature and melter operations due to the minimal ~0.2 weight percent (wt%) increase in Gd concentration relative to previous sludge batches [Gap closed for SB11]; 5) As observed previously, Gd is expected to enter the off-gas system via physical entrainment, but at a slightly higher concentration than what was observed for SB9 melter off-gas pluggage deposits (0.07 wt%) [Gap closed for SB11] ; 6) There are no expected impacts on DWPF recycle or the Recycle Collection Tank glycolate destruction process. [Gap closed for SB11]; 7) Gd is projected to be a trace component in the SB11 glass (<0.5 wt%) and can be ignored for process control. Trace components do not significantly impact glass durability, thus the conclusions of the previous Product Consistency Test evaluation at a fissile mass loading of 2,500 g fissile/m3 glass still applies to SB11. The ~0.1 wt% increase in Gd2O3 concentration relative to the previous study will not impact the predictability of SB11 glass with the DWPF Product Composition Control System (PCCS) models for durability or the acceptability of glass according to the Waste Acceptance Product Specifications (WAPS) criterion for product consistency [Gap closed for SB11]; 8) No additional Toxicity Characteristic Leaching Procedure testing is necessary for SB11 and the hazardous waste specification of the SB11 DWPF waste form is unchanged after the addition of the ABD stream [Gap closed for SB11]. The following summarizes the evaluation of the impacts of adding two-thirds of the ABD material to Tank 51 prior to LTAD: 1) The addition of two-thirds of the ABD increases overall aluminum mass from 1.39×10 4 kg to 1.64×10 4 kg (15.5% ABD Al). The form of the insoluble portion of the Al resulting from ABD addition should be the more readily dissolved Al(OH) 3 and amorphous forms. The portion of the ABD aluminum that is processed by LTAD is expected to be completely soluble, thus requiring that less of the boehmite in the sludge be dissolved to reach the same Al target in the SB. [Gap closed for SB11]; The expected LTAD impact on other components, as related primarily to the components in ABD, are discussed. Gd is expected to remain insoluble during LTAD and not impact the solubility of other components. [Gap closed for SB11]; The addition of two-thirds of the ABD increases overall projected SB11 uranium mass from 4,740 kg to 13,100 kg (63% ABD U) and the projected plutonium mass from 86.0 kg to 89.5 kg (3.9% ABD Pu). The addition of all of the ABD increases overall projected SB11 uranium mass from 4,740 kg to 16,100 kg (70% ABD U) and the projected plutonium mass from 86.0 kg to 90.4 kg (5.3% ABD Pu). The 235 U enrichment will be ≤5%. The fissile uranium will be adequately poisoned by Gd and the fissile Pu will be adequately poisoned by Fe from the sludge. [Gap closed for SB11]; There is a low risk that ABD addition will impact the rheology or pumpability of the slurry. There is a low but higher risk of ABD addition prior to LTAD impacting the settling rate; Based on the evaluation of adding two-thirds of the ABD material and all of the ABD material prior to the LTAD process, there is no volume or mass limit that would need to be imposed on ABD additions prior to LTAD. [Gap closed for SB11]. Revision 1 of this report addresses a variation on the ABD additions and LTAD strategy where sodium hydroxide additions for LTAD may be performed intermittently or concurrently with an ABD addition window. The proposed change does not alter the conclusions of this evaluation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗