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DOE OSTI2012

Mercury Reduction and Removal from High Level Waste at the Defense Waste Processing Facility - 12511

The Defense Waste Processing Facility processes legacy nuclear waste generated at the Savannah River Site during production of enriched uranium and plutonium required by the Cold War. The nuclear waste is first treated via a complex sequence of controlled chemical reactions and then vitrified into a borosilicate glass form and poured into stainless steel canisters. Converting the nuclear waste into borosilicate glass is a safe, effective way to reduce the volume of the waste and stabilize the radionuclides. One of the constituents in the nuclear waste is mercury, which is present because it served as a catalyst in the dissolution of uranium-aluminum alloy fuel rods. At high temperatures mercury is corrosive to off-gas equipment, this poses a major challenge to the overall vitrification process in separating mercury from the waste stream prior to feeding the high temperature melter. Mercury is currently removed during the chemical process via formic acid reduction followed by steam stripping, which allows elemental mercury to be evaporated with the water vapor generated during boiling. The vapors are then condensed and sent to a hold tank where mercury coalesces and is recovered in the tank's sump via gravity settling. Next, mercury is transferred from the tank sump to a purification cell where it is washed with water and nitric acid and removed from the facility. Throughout the chemical processing cell, compounds of mercury exist in the sludge, condensate, and off-gas; all of which present unique challenges. Mercury removal from sludge waste being fed to the DWPF melter is required to avoid exhausting it to the environment or any negative impacts to the Melter Off-Gas system. The mercury concentration must be reduced to a level of 0.8 wt% or less before being introduced to the melter. Even though this is being successfully accomplished, the material balances accounting for incoming and collected mercury are not equal. In addition, mercury has not been effectively purified and collected in the Mercury Purification Cell (MPC) since 2008. A significant cleaning campaign aims to bring the MPC back up to facility housekeeping standards. Two significant investigations are being undertaken to restore mercury collection. The SMECT mercury pump has been removed from the tank and will be functionally tested. Also, research is being conducted by the Savannah River National Laboratory to determine the effects of antifoam addition on the behavior of mercury. These path forward items will help us better understand what is occurring in the mercury collection system and ultimately lead to an improved DWPF production rate and mercury recovery rate. (authors)

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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 OSTI2018

Toward an Understanding of Surface Layer Formation, Growth, and Transformation at the Glass–Fluid Interface

Silicate glass is a metastable and durable solid that has application to a number of energy and environmental challenges (e.g., microelectronics, fiber optics, and nuclear waste storage). If allowed to react with water over time silicate glass develops an altered layer at the solid-fluid interface. In this study, we used borosilicate glass (LAWB45) as model material to develop a robust understanding of altered layer formation (i.e., amorphous hydrated surface layer and crystalline reaction products). Experiments were conducted at high surface area-to-volume ratio (~200,000 m-1) and 90 °C in the pressurized unsaturated flow (PUF) apparatus for 1.5-years to facilitate the formation of thick altered layers and allow for the effluent solution chemistry to be monitored continuously. A variety of microscopy techniques were used to characterized reacted grains and suggest the average altered layer thickness is 13.2 ± 8.3 μm with the hydrated and clay layer representing 74.8% and 25.2% of the total altered layer, respectively. This estimate is within the experimental error of the value estimated from the B release rate data (~10 ±1 μm/yr) over the 1.5-year duration. PeakForce® quantitative nanomechanical mapping results suggest the hydrated layer has a modulus that ranges between ~20 to 40 GPa, which is in the range of porous silica that contains from ~20 to ~50% porosity, yet significantly lower than dense silica (~70 to 80 GPa). Scanning transmission electron microscopy (STEM) images confirm the presence of pores and an analysis of a higher resolution image of a region provides a qualitative estimate of ≥ 22% porosity in this layer with variations in the hydrated layer in void volume with increasing distance from the unaltered glass. Chemical composition analyses, based on a combination of time-of-flight secondary-ion mass spectrometry (ToF-SIMS) and scanning electron microscopy with X-ray energy dispersive spectroscopy (EDS) and STEM-EDS, clearly show that the altered layer is mainly composed of Al, H, Si, and O with the clay layer being enriched in Li, Zn, Fe, and Mg. The amorphous hydrated layer is enriched in Ca, H, and Zr with a minor amount of K. Furthermore, ToF-SIMS results also suggest the B profile is anti-correlated with the H profile in the hydrated layer. Our selected-area electron diffraction results suggest the structure of the hydrated layer closely resembles opal-AG (amorphous gel-like) with an average crystallite size of ~0.7 nm which is smaller than the critical nucleus for silica nanoparticles (i.e., 1.4 to 3 nm). These results suggest the hydrated layer is more consistent with a polymeric gel rather than a colloidal gel and is comprised of molecular units (<1 nm in size) that result from the difficult to hydrolyze bonds, such as Si—O—Zr units, during the glass corrosion process. The size of individual particles or molecular units is a function of formation conditions (e.g., pH, ionic strength, nano-confinement, solute composition) in the hydrated layer.

36 MATERIALS SCIENCE↗

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.

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