Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “glass dissolution”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 55 records · Page 3

Radiation effects on the structure and alteration behavior of an SiO 2 –Al 2 O 3 –B 2 O 3 –Na 2 O glass

As borosilicate glasses are used in many countries to immobilize fission products and minor actinides after spent fuel reprocessing before storage in a deep geological repository, assessing that their chemical durability is of paramount importance. Here, pristine and preirradiated (952 MeV, 136Xe) SiO 2 –B2O 3 –Al2O 3 –Na 2 O glasses with the same molar ratios as in the French SON68 and ISG glasses have been subjected to aqueous corrosion in deionized water and in silica-saturated solution to measure the initial and longer term alteration rates. Pristine and preirradiated glasses corrode following the same mechanisms, but the preirradiation has a strong impact on the initial dissolution rate (increase by a factor of 5.6), and on the alteration layer depth in silica-saturated conditions (by two- to threefolds). The later result is related to the formation of a more porous, less passivating gel on the preirradiated glass specimen. Using both experimental spectroscopies (NMR, IR, and SFG) and classical molecular dynamics, the radiation effects on the glass structure and water diffusion have been assessed. After preirradiation, the density and the polymerization degree of the glass decrease, whereas the topological disorder increases. In consequence, water diffusion accelerates. These observations allow to correlate the radiation impact on the alteration behavior to the structural changes.

36 MATERIALS SCIENCE↗

Effect of glass forming additives on low-activity waste feed conversion to glass

A significant effort was invested in the past to develop and refine mathematical models that relate the composition of nuclear waste glasses with their properties, such as viscosity, electrical conductivity, or chemical durability. However, less attention has been paid to the formulation of the melter feed itself, such as the chemical form and the particle size of the glass forming and modifying additives (GFMA), which have a significant effect on the feed-to-glass conversion process during melting. To address this issue, we systematically changed the mineral composition of a simulated low-activity waste melter feed and inspected its melting behavior. When substituting minerals with corresponding oxides and hydroxides, we found that different alumina sources (kyanite, gibbsite, boehmite, or corundum) had the strongest effect on the feed melting process, whereas the sources of Ca, Mg, and Zr had little effect. Further, the x-ray diffraction analysis showed that the alumina sources differ in their dissolution kinetics: early dissolving alumina sources, such as gibbsite (Al(OH) 3 ) and boehmite (AlO(OH)), increase the transient glass-forming melt viscosity at early stages, when gases still evolve, causing extended foaming, whereas alumina sources that dissolve at high temperatures, such as kyanite (Al 2 SiO 5 ) and corundum (Al 2 O 3 ), keep the transient glass-forming melt viscosity low and lead to a faster foam collapse. Using viscosity-composition relationship to estimate the viscosity of transient glass-forming melts in the primary foaming range, we found that the primary foam began to collapse at 360 to 800 Pa s, and fully collapsed between 65 and 260 Pa s. This result agrees with our previous studies, according to which the glass-forming melt viscosity at the cold cap foam bottom ranged from 24 to 85 Pa s.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Determining the Solubility Behavior of Kogarkoite in Simulated Nuclear Waste

Kogarkoite (Na 3 FSO 4 ) is a sparingly soluble fluoride–sulfate double salt that has been identified in high level nuclear waste sludge at the Hanford Site and, more recently, in sludge batch compilation samples at the Savannah River Site (SRS). Due to its complex dissolution behavior, which exhibits an inverse dependence on sodium ion activity, the presence of this mineral poses significant challenges to waste retrieval and processing. Incomplete dissolution during sludge washing can lead to the retention of fluoride and sulfate in the high-level waste feed, potentially causing the formation of corrosive, immiscible molten salt layers, known as "glass gall,” in vitrification melters. Current efforts to optimize flowsheet parameters and wash-water volumes are hindered by the absence of a commercially available, certified reference material, which prevents the accurate calibration of analytical methods and the verification of dissolution kinetics. To address this critical gap, this research focuses on the laboratory synthesis of pure Kogarkoite to serve as a standard for comprehensive solubility and washing performance testing. A coupled synthesis and simulant campaign was executed using an evaporative crystallization protocol designed to replicate the dynamic concentration effects observed in tank farm operations. Thirteen simulant matrices were prepared by dissolving systematically varied ratios of sodium fluoride (NaF) and sodium sulfate (Na 2 SO 4 ) in deionized water under three distinct caustic regimes: 0.0 g (control), 4.0 g (~1 M), and 12.0 g (~3 M) sodium hydroxide (NaOH). While thermodynamic equilibrium models suggest that high-caustic environments should favor the stability of the double salt7, results from this evaporative study at 25 0 C revealed a distinct kinetic divergence. Simulants with high hydroxide loading predominantly yielded large, blocky crystals of sodium sulfate decahydrate (Na 2 SO 4 .10H 2 O). Successful synthesis of pure Kogarkoite was achieved exclusively in specific NaOH-free compositional windows, where the precipitate manifested as fine, opaque granular aggregates. Ion chromatography (IC) analysis confirmed phase purity through the simultaneous stoichiometric depletion of both fluoride and sulfate from the supernatant. This successful synthesis establishes a reproducible route to generate bulk Kogarkoite, enabling the subsequent phase of quantitative dissolution testing using inhibited water to optimize sludge-batch assembly.

Sarker, Md Sharif [Florida International Univ. (FI↗

Corrosion behavior of Monofrax K-3 refractory in borosilicate-based model low activity waste glass melts

Owing to its good chemical and thermal durabilities at high temperatures, Monofrax K-3 refractory is widely used in nuclear waste vitrification as a lining material in melting vessels. However, the corrosion of K-3 refractory during the vitrification of nuclear waste is a serious problem because it affects the melter's safety, performance, and lifetime. Therefore, in the present study, we have focused on unearthing the impact of glass network formers, such as SiO 2 , B 2 O 3 , and Al 2 O 3 , in a model nuclear waste glass composition on the corrosion of Monofrax K-3 refractory. The corrosion tests have been performed per ASTM C621 at 1150°C for 5 days. The dimensional measurements on corroded K-3 refractory suggest that Al 2 O 3 and SiO 2 tend to reduce the refractory corrosion (neck loss), with the effect of Al 2 O 3 being significant. A corroded region on the K-3 refractory at the melt–refractory interface is observed. The corrosion occurs via a coupling of the melt infiltration induced by a capillary effect and the dissolution of Al, Mg, and Fe components from K-3 into the melt through chemical reactions. A Cr-rich layer is retained on the glass contact surface of the corroded K-3 refractory.

36 MATERIALS SCIENCE↗

Sulfur speciation in dacitic melts using X-ray absorption near-edge structure spectroscopy of the S K -edge (S-XANES): Consideration of radiation-induced changes and the implications for sulfur in natural arc systems

Abstract The synchrotron technique of micro X-ray absorption near-edge structure spectroscopy at the sulfur K-edge (S-XANES) provides a unique opportunity to measure the proportion of different oxidation states of sulfur (S) in silicate glasses. Although applied extensively in the analysis of basaltic silicate glasses, few S-XANES studies have investigated variations in S oxidation states with fO2 in felsic silicate glasses. In addition, no study has systematically compared the S-XANES results obtained from the same samples at different photon flux densities to quantify the relationship between exposure time and changes in S speciation in silicate glass, as has been done for Fe and V. This study evaluates observed differences in S speciation measured in experimentally produced H2O-saturated dacitic glasses over a range of reducing to oxidizing conditions (from log fO2 = ΔFMQ-0.7 to ΔFMQ+3.3; FMQ is the fayalite-magnetite-quartz mineral redox buffer) and equilibrated at 1000 °C and 300 MPa. S-XANES spectra were collected at three different photon flux densities using three microspectroscopy beamlines. As is observed in S-XANES analyses of basaltic silicate glasses, beam-induced changes to the S6+/ΣS are observed as a function of photon flux density and beam exposure time. Our results demonstrate that silicate glasses of dacitic composition undergo beam-induced photo-reduction in samples equilibrated at ΔFMQ > +1.75 and photo-oxidation if equilibrated at ΔFMQ < +1. The time required to observe beam-induced changes in the spectra varies as a function of flux density, and our study establishes an upper photon density limit at ~1.0 × 1012 photons/µm2. The S6+/ΣS calculated from spectra collected below this absorbed photon limit at intermediate flux densities (~1–4 × 109 photons/s per µm2) are affected by beam damage, as no conditions were found to be completely free of beam-induced changes. However, the S6+/ΣS ratios calculated below the limit at intermediate flux densities are consistent with thermodynamic constraints, demonstrating that S6+/ΣS ratios calculated from S-XANES spectra can be considered reliable for estimating the oxygen fugacity. Our results carry important implications for the S budget of felsic magmas and dissolution mechanisms in evolved melts. While our results from all three flux densities show the presence of S4+ dissolved in relatively oxidized (ΔFMQ > +1.75) dacitic glass, even in the spectra exposed to the lowest photon densities, we are unable to rule out the possibility that the S4+ signal is the result of instantaneous X-ray irradiation induced beam damage using S-XANES alone. When our spectra are compared to S-XANES spectra from basaltic silicate glasses, important differences exist in the solubility of S2– and S6+ between dacitic silicate melts, pointing to differences in solubility mechanisms as melt composition changes. This study highlights the need for further investigation into beam damage systematics, presence of S4+, and the solubility mechanisms of different oxidation states of S as silicate melt composition changes.

Geochemistry & Geophysics↗

Dynamic and static corrosion of chromium-alumina refractory in simulated nuclear waste glass

Monofrax® K-3, a chromium–alumina refractory, is widely used in nuclear waste glass melters due to its high durability. However, aggressive vitrification conditions still lead to corrosion that limits melter lifespan. Here, this study investigates two key corrosion modes, subsurface and melt-line corrosion, under static and dynamic conditions, using a custom setup enabling precise control of melt velocity and temperature. Experimental results, interpreted using diffusion-limited dissolution models, show that (i) subsurface corrosion increases by approximately 10% per mm/s increase in melt velocity, while melt-line corrosion remains virtually unaffected, (ii) corrosion rate dependence on temperature is inversely proportional to melt viscosity, and (iii) glass composition affects both corrosion modes similarly, based on Cr 2 O 3 , Al 2 O 3 , and alkali contents. The experimental data also allowed the determination of diffusion coefficients of major dissolving species, which were found to be consistent between the melt-line and subsurface corrosion models, indicating their suitability for future CFD studies.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗

Flax–Reinforced Vitrimer Epoxy Composites Produced via RTM

Composite laminates were produced by RTM using similar glass and flax fabrics and both vitrimer epoxy and aerospace-grade epoxy, both formulated for liquid molding. Tensile and flexural properties were measured and compared, revealing that the vitrimer composites exhibited equivalent performance in flexural strength and tensile modulus, but slightly lower performance in tensile strength relative to reference epoxy composites. In general, glass–fiber composites outperformed flax–fiber composites in tension. However, both glass and flax–fiber composites yielded roughly equivalent flexural strength and tensile modulus-to-weight ratios. Flax fabrics were recovered from composites by matrix dissolution, and a second-life laminate showed full retention of the mechanical properties relative to those produced from fresh flax. Finally, a demonstration of re-forming was undertaken, showing that simple press-forming can be used to modify the composite shape. However, re-forming to a flat configuration resulted in local fiber damage and a decrease in mechanical properties. An alternative forming method was demonstrated that resulted in less fiber damage, indicating that further refinements might lead to a viable forming and re-forming process.

Martinez, Patricio (ORCID:0000000225884389)↗

Tank 11H Low Temperature Aluminum Dissolution and Inhalation Dose Potential Analyses at Savannah River Site – 26018

Currently, there is approximately 34 million gallons of high-level radioactive tank waste in the Tank Farm at the Savannah River Site (SRS). The ultimate goal of operations at the Tank Farm is to remove the high level waste (HLW) from the tanks followed by stabilization of the waste through vitrification of the HLW into glass or grouting the decontaminated waste into saltstone. After bulk removal of the HLW consisting of sludge, saltcake, and supernatant, further efforts are made to reduce the residual waste present in the tank in order to declare preliminary cease waste removal (PCWR) signifying completion of HLW removal. These reduction efforts can include tank washing to remove soluble salts and radioisotopes and dissolution of solids including aluminum. Aluminum in the form of gibbsite and boehmite is relatively insoluble in water. Through addition of aqueous sodium hydroxide, the aluminum can be dissolved at mild temperatures. In order for the waste tank to meet closure mode requirements of the Concentration, Storage, and Transfer Facilities (CSTF), which includes the Tank Farm, Documented Safety Analysis (DSA), a component of the safety basis, the inhalation dose potential (IDP) and the radiolytic hydrogen generation rate of the stored waste must be demonstrated to be lower than their respective designated limits. These parameters are calculated from measured radiochemical analyses of isotopes that emit a high amount of radioactivity including Cs-137, Sr-90, Pu-238, Pu-239, Pu-240, Pu-241, Am-241, and Cm-244. Following the low temperature aluminum dissolution (LTAD) process, Tank 11H slurry samples were pulled from the tank and sent to Savannah River National Laboratory (SRNL) to measure the extent of aluminum dissolution, hydroxide concentration, densities of slurry and supernatant, weight percent solids analyses, and radionuclide activities. The analyses of the composite sample found that approximately 90% of the total aluminum in the slurry was dissolved, indicating successful reduction of the insoluble aluminum in the waste tank. Additionally, the weight percent insoluble solids (slurry basis) measurement of the composite sample was found to be approximately 1%, demonstrating that minimal solids still remain in the tank. Finally, the radiochemical analyses of the composite sample determined that the waste contents of the tank met the IDP and radiolytic hydrogen generation rate requirements of the CSTF DSA. These measurements have shown that the LTAD process in Tank 11H was successful in waste reduction efforts and a positive step towards declaring PCWR and tank closure at SRS.

Dekarske, John [Savannah River National Laboratory↗

Impacts of Fast Critical Assembly Fuel Discards on Liquid Waste Processes

The Savannah River Nuclear Solutions (SRNS) Fast Critical Assembly (FCA) mission is reestablishing the electrolytic dissolver for processing of Pu and Pu-U materials clad in stainless steel (SS). H-Canyon is planning to dissolve and neutralize FCA fuel without recovering the special nuclear material (i.e., Pu) prior to discarding to the Concentration, Storage, and Transfer Facilities (CSTF) operated by the Savannah River Mission Completion (SRMC) Liquid Waste (LW) Organization. The FCA discards will be combined with sludge in the CSTF after Low Temperature Aluminum Dissolution (LTAD) if needed. The combined waste will be washed, concentrated, and vitrified at the Defense Waste Processing Facility (DWPF). The high level waste canisters produced will be temporarily stored in the Glass Waste Storage Buildings prior to transfer to a future federal repository. Decants from LTAD and sludge washing will be combined with DWPF recycle and dissolved salt cake to prepare salt batches for processing in the Salt Waste Processing Facility. The resulting decontaminated salt solution will be processed in the Saltstone Production Facility.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Phosphate and Fluoride Processing Options for Hanford Sludge

The Direct Feed High-Level Waste (DFHLW) strategy represents an alternative flowsheet to bypass the Hanford Waste Treatment and Immobilization Plant (WTP) Pretreatment Facility. Key processing functions previously captured during baseline pretreatment operations include leaching and washing prior to solids concentration and should be considered in the new DFHLW flowsheet to maximize waste feed loading, minimize high-level waste (HLW) volume, and mitigate corrosion challenges associated with vitrification of high phosphate and fluoride concentrations. The effectiveness and efficiency of sludge washing can have a substantial impact on DST space and mission duration. Two target species that benefit significantly from washing are phosphate and fluoride. Commonly present in many high-level wastes at Hanford, phosphate and fluoride are found predominately in the form of salt precipitates and can have adverse and detrimental effects on glass waste loading. Additionally, fluoride produces melter off-gas that creates corrosion risks in the off-gas system piping. Determining the solubilities of fluoride and phosphate salts, as well as the dissolution kinetics under potential HLW processing conditions will help to mitigate future operational risks and mature flowsheet technical bases.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Cs 3 Bi 2 I 9 -hydroxyapatite composite waste forms for cesium and iodine immobilization

Perovskite-based ceramic composites were developed as potential waste form materials for immobilizing cesium (Cs) and iodine (I) with high waste loadings and chemical durability. The perovskite Cs 3 Bi 2 I 9 has high Cs (22 wt%) and I (58 wt%) content, and thus can be used as a potential host phase to immobilize these critical radionuclides. In this work, the perovskite Cs 3 Bi 2 I 9 phase was synthesized by a cost effective solution-based approach, and was embedded into a highly durable hydroxyapatite matrix by spark plasma sintering to form dense ceramic composite waste forms. The chemical durabilities of the monolithic Cs 3 Bi 2 I 9 and Cs 3 Bi 2 I 9 -hydroxyapatite composite pellets were investigated by static and semi-dynamic leaching tests, respectively. Cs and I are incongruently released from the matrix for both pure Cs 3 Bi 2 I 9 and composite structures. The normalized Cs release rate is faster than that of I, which can be explained by the difference in the strengths between Cs-I and Bi-I bonds as well as the formation of insoluble micrometer-sized BiOI precipitates. The activation energies of elemental releases based on dissolution and diffusion-controlled mechanisms are determined with significantly higher energy barriers for dissolution from the composite versus that of the monolithic Cs 3 Bi 2 I 9 . The ceramic-based composite waste forms exhibit excellent chemical durabilities and waste loadings, commensurate with the state-of-the-art glass-bonded perovskite composites for I and Cs immobilization.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Modeling glass degradation and release of radionuclides from vitrified waste for performance assessment simulations

The release of radionuclides initially encapsulated in a slowly degrading solid waste form and contained in an eventually corroding canister defines the source term for numerical simulations for the assessment of a geologic repository for high-level radioactive waste. While the details of waste degradation, canister corrosion, and dissolution and mobilization of the radionuclides in pore water include complex chemical reaction and transport processes that are coupled to the thermal, hydrological, microbiological, and mechanical conditions in the repository, the source-term model suitable for use in a numerical performance assessment model should be a defensible abstraction of these mechanisms. We developed a radiological source-term model and implemented it into a non-isothermal flow and transport simulator. While the proposed source-term model is applicable to various waste forms, canister systems, and disposal concepts, we specifically considered radionuclide releases from vitrified high-level waste placed in a cylindrical canister disposed in a deep vertical borehole repository. In this model, waste degradation is a function of temperature, and it can be adjusted to evaluate the influence of and propagate uncertainties in pH, passivation reactions, and chemical conditions as well as geometrical factors. The time-dependent, congruent release of safety-relevant radionuclides present in the decaying inventory is then calculated. Finally, the radionuclides are mobilized by diffusive and advective transport according to the thermo-hydraulic conditions prevailing in the near field of the repository, from where they migrate through the geosphere to the accessible environment. We examine the influence of the source-term model’s parameters on performance assessment calculations through sensitivity and uncertainty propagation analyses, identifying influential factors and confirming the upper bound of their impact. These considerations align with the overarching goal of repository design, which is to demonstrate that engineered and natural barriers can collectively delay radionuclide migration for timescales far exceeding human planning, thereby providing multiple, redundant barriers against environmental contamination.

iTOUGH2↗

Realizing high-capacity all-solid-state lithium-sulfur batteries using a low-density inorganic solid-state electrolyte

Abstract Lithium-sulfur all-solid-state batteries using inorganic solid-state electrolytes are considered promising electrochemical energy storage technologies. However, developing positive electrodes with high sulfur content, adequate sulfur utilization, and high mass loading is challenging. Here, to address these concerns, we propose using a liquid-phase-synthesized Li 3 PS 4 -2LiBH 4 glass-ceramic solid electrolyte with a low density (1.491 g cm −3 ), small primary particle size (~500 nm) and bulk ionic conductivity of 6.0 mS cm −1 at 25 °C for fabricating lithium-sulfur all-solid-state batteries. When tested in a Swagelok cell configuration with a Li-In negative electrode and a 60 wt% S positive electrode applying an average stack pressure of ~55 MPa, the all-solid-state battery delivered a high discharge capacity of about 1144.6 mAh g −1 at 167.5 mA g −1 and 60 °C. We further demonstrate that the use of the low-density solid electrolyte increases the electrolyte volume ratio in the cathode, reduces inactive bulky sulfur, and improves the content uniformity of the sulfur-based positive electrode, thus providing sufficient ion conduction pathways for battery performance improvement.

25 ENERGY STORAGE↗

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↗

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↗

Viscosity of transient glass-forming melt and its relation to foaming during batch-to-glass conversion

Primary foam, which affects the heat transfer into the glass batch and the final glass quality, occurs when a sufficient quantity of transient glass-forming melt evolves with viscosity low enough to close the open porosity of the reacting glass batch. To better understand how the fraction of transient melt and its viscosity affect the primary foam temperature range, we determined, with x-ray diffraction, the fraction and composition of the transient glass-forming melt in a heated waste glass melter feed as a function of temperature. Then we prepared a set of transient melts that occurred within the foaming temperature interval and measured their viscosities with spindle and falling sphere viscometers. Further, the results agree with the Adam-Gibbs and VFT viscosity-composition models, even for transient melt compositions outside of the models compositional validity range. As silica and other refractory particles dissolved in the predominantly borate transient melt while temperature increased, viscosity increased from the initial value of ~500 Pa s at the onset of foaming (~650°C) to a maximum of ~770 Pa s when silica dissolution was almost complete (~700°C). As temperature increased further, transient melt viscosity decreased to ~220 Pa s (~850°C) when the primary foam collapsed.

36 MATERIALS SCIENCE↗

Reassessment of Dose Contributions from the M-Area Glass Special Waste Form Buried in the E-Area Low-Level Waste Facility

During a recent revision to the groundwater (GW) and inadvertent human intruder (IHI) screening analysis report (Aleman and Hamm, 2023), weaknesses within the screening logic were observed resulting in U-235 being brought back into the list of radionuclides requiring inventory limits within the M-Area Glass SWF. In addition, while reviewing the earlier vadose zone (VZ) transport analysis efforts, an input deck error was observed in the U-234G VZ transport simulation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Thermally Insulating, Thermal Shock Resistant Calcium Aluminate Phosphate Cement Composites for Reservoir Thermal Energy Storage

This paper presents the use of hydrophobic silica aerogel (HSA) and hydrophilic fly ash cenosphere (FCS) aggregates for improvements in the thermal insulating and mechanical properties of 100- and 250 °C-autoclaved calcium aluminate phosphate (CaP) cement composites reinforced with micro-glass (MGF) and micro-carbon (MCF) fibers for deployment in medium- (100 °C) and high-temperature (250 °C) reservoir thermal energy storage systems. The following six factors were assessed: (1) Hydrothermal stability of HSA; (2) Pozzolanic activity of the two aggregates and MGF in an alkali cement environment; (3) CaP cement slurry heat release during hydration and chemical reactions; (4) Composite phase compositions and phase transitions; (5) Mechanical behavior; (6) Thermal shock (TS) resistance at temperature gradients of 150 and 225 °C. The results showed that hydrophobic trimethylsilyl groups in trimethylsiloxy-linked silica aerogel structure were susceptible to hydrothermal degradation at 250 °C. This degradation was followed by pozzolanic reactions (PR) of HSA, its dissolution, and the formation of a porous microstructure that caused a major loss in the compressive strength of the composites at 250 °C. The pozzolanic activities of FCS and MGF were moderate, and they offered improved interfacial bonding at cement-FCS and cement-MGF joints through a bridging effect by PR products. Despite the PR of MGF, both MGF and MCF played an essential role in minimizing the considerable losses in compressive strength, particularly in toughness, engendered by incorporating weak HSA. As a result, a FCS/HSA ratio of 90/10 in the CaP composite system was identified as the most effective hybrid insulating aggregate composition, with a persistent compressive strength of more than 7 MPa after three TS tests at a 150 °C temperature gradient. This composite displayed thermal conductivity of 0.28 and 0.35 W/mK after TS with 225 and 150 °C thermal gradients, respectively. These values, below the TC of water (TC water = 0.6 W/mK), were measured under water-saturated conditions for applications in underground reservoirs. However, considering the hydrothermal disintegration of HSA at 250 °C, these CaP composites have potential applications for use in thermally insulating, thermal shock-resistant well cement in a mid-temperature range (100 to 175 °C) reservoir thermal energy storage system.

58 GEOSCIENCES↗