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Wastewater Brine Purification and Recovery through Electrodialysis Ion Exchange

Reutilizing resources onboard the International Space Station (ISS) and for future deep space missions are critical for mission longevity and sustainability. Wastewater brine produced from water recovery systems contain chemical species that could be processed into a potential fertilizer for future plant systems. This can be achieved through a process called electrodialysis ion exchange. Wastewater containing inorganic salt components are fed through a series of ion exchange membranes to produce fertilizer (a phosphate rich stream), electrolysis-grade water, and other useful commodities. Electrodialysis cells consisting of anion and cation exchange membranes, monovalent anion exchange membranes, and bipolar membranes were utilized to achieve selective ion exchange. The use of the electrodialysis cells were effective for both water extraction and ion separation. Ions successfully diffused across their respective membranes into the concentrate, acid, and base streams. This resulted in pure water, a phosphate rich stream, and a separate anion/hydrogen and cation/hydroxide stream. However, sulfate and some phosphate ions were able to diffuse through the monovalent anion exchange membrane into the acid stream. This resulted in predominantly phosphate ions remaining in the concentrate. Optimization of the process was accomplished by altering flowrates of each stream and initial volumes, adjusting the power input and resulting current through each cell, and varying the starting parameters by splitting the inorganic waste input into the diluate and the concentrate. As expected, increasing the flowrate and the power input to each cell reduced the overall time of the process. However, mission constraints may require a longer duration process in order to reduce the power consumption. Further analysis will be required to determine the power input necessary to achieve ion diffusion effectively and in a timely manner.

Ion Exchange Membranes↗

Crystalline Aluminosilicates for Radionuclide Remediation

Crystalline aluminosilicates have a wide range of applications such as media for the capture and/or immobilization of radionuclides. Several different types of metal-exchanged zeolites have been successfully demonstrated as effective sorbents for gaseous, aqueous, or solid-state capture and immobilization of radionuclides including iodine [e.g., I2(g), CH3I(g)], xenon, krypton, radium (e.g., 226Ra and 228Ra), selenium (e.g., 79Se6+), technetium (e.g., 99TcO4-), strontium (e.g., 90Sr2+), cesium (e.g., 135Cs+, 137Cs+), mixed alkali halide streams (i.e., iodide and chloride salts), and actinides (e.g., U6+, Th4+). These zeolites include sodalite (SOD), mordenite (MOR or Z-type), faujasite (FAU), Linde type-A (LTA), cancrinite (CAN), gismondine (GIS), and heulandite (or clinoptilolite) (HEU). The structural building blocks of these seven materials vary from cages (e.g., a-cages, ß-cages, e-cages) to channels linking various cage types and these are mostly comprised of aluminosilicates. This chapter will delve into the crystallographic differences between these framework materials and provide a brief overview of literature data where these seven crystalline aluminosilicates have been utilized for the capture and/or immobilization of radionuclides. Metal-exchanged zeolites (e.g., MOR, FAU, LTA) have often been the source of study for gaseous capture of iodine where the more promising compositions include Ag-exchanged sorbents; these types of materials can be made in a variety of robust forms including berl saddles and extrudates. Sodalite can be used as an immobilization strategy for halide waste streams (e.g., Cl, I), including those from electrochemical reprocessing of used nuclear fuel, as well as TcO4–. Cancrinite has been demonstrated to enhance sorption capacity of sediments for several problematic ionic radionuclides including Cs+, SeO42–, and TcO4–. Actinides such as U6+ and Th4+ have been successfully captured from solutions using MOR, FAU, LTA, and HEU-type aluminosilicates. Finally, mixtures of aluminosilicates have been included in a set of zeolite materials that show promise as sorbents for 226Ra and 228Ra removal from mine waters.

zeolites, aluminosilicates, radionuclide remediati↗

Studies on the Encapsulation of Brine Generated from a Process Using Selective Electrodialysis Reversal

This topical report summarizes the laboratory testing, geochemical modeling, and mineralogical testing conducted to examine the efficacy of wastewater encapsulation for the brine generated from the Saltworks Flex electrodialysis reversal technology. Within the laboratory testing, the short- and long-term physical properties of seven separate mix designs were examined along with the leaching of each mix design. Additionally, geochemical thermodynamic modeling was used along with mineralogical analyses to help gain a fundamental understanding of the chemistry involved in the encapsulation process. Encapsulation is a process that results in the immobilization of constituents of interest by creating an engineered material that contains the waste within its matrix, both physically and chemically, in such a manner that the receiving environment is protected from adverse effects. Within the context of this study, encapsulation mix designs involved mixing the Saltworks brine with fly ash and various additives, principally lime (calcium oxide) and cement, to form a paste that eventually sets and hardens into a monolith upon placement in a landfill. Moisture conditioning fly ash with wastewater sources for on-site landfill disposal is a process that is in use at some sites and has been presented as a viable option to dispose of liquid wastes. However, because chemical reactions that can lead to sequestration of constituents of interest are not optimized, leaching can occur. Therefore, in this study, mixes were created to mimic moisture conditioning, as well as engineered pastes of varying mix designs. The study found that engineered encapsulation mixes provide significant resistance to leaching compared to moisture conditioning alone. The best performing encapsulation mix used lime as the additive, which proved superior to the use of cement. After 90 days of curing, this encapsulated mix was 5.5 times stronger than the moisture conditioned mix. Additionally, this mix exhibited an approximate 2.5 order of magnitude drop in hydraulic conductivity compared to the moisture conditioned mix. When comparing leaching, nearly all the chloride was found to evolve from the moisture conditioned mix, whereas only 6.7% of the chlorine contained in the best performing encapsulated mix leached out. Mineralogical analysis via X-ray diffraction (XRD) showed a significant concentration of the chloride sequestering mineral, Friedel’s salt, in the best performing encapsulation mix, whereas none was observed in the moisture conditioned mix, pointing to it being the primary chemical mechanism for chloride retention, along with the physical impermeability of the material. With respect to the modeling study, the model was successful in predicting the presence of most mineral phases observed in the XRD analysis.

01 COAL, LIGNITE, AND PEAT↗

Mercury Dispersion Modeling and Purge Ventilation Stack Height Determination for Tank 37H

The Savannah River National Laboratory (SRNL) Atmospheric Technologies Group performed atmospheric dispersion modeling and analysis of mercury emissions from the H-Tank Farm - Tank 37 ventilation system exhaust to evaluate whether the Short-Term Exposure Limit (STEL) for mercury is exceeded during salt dissolution activities. This analysis was also used to establish a minimum stack height at which ambient mercury concentration would not exceed this regulatory limit. The American Meteorological Society/Environmental Protection Agency Regulatory Model (AERMOD) was used as the dispersion modeling tool for this analysis. Results indicate that the default 10-foot stack results in ground level concentrations that significantly exceed the standards for the mercury discharge scenarios evaluated. A 35-foot stack was necessary to raise the exhaust plume centerline from Tank 37H to a height that prevents mercury exposure problems for the three discharge concentrations modeled.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Development and Optimization of Voltammetric Methods for Real Time Analysis of Electrorefiner Salt with High Concentrations of Actinides and Fission Products

This project addresses the problem of achieving accurate material control and accountability (MC&A) around pyroprocessing electrorefiner systems. Spent nuclear fuel pyroprocessing poses a unique challenge with respect to reprocessing technology in that the fuel is never fully dissolved in the process fluid. In this case, the process fluid is molten, anhydrous LiCl-KCl salt. Therefore, there is no traditional input accountability tank. However, electrorefiners (ER) accumulate very large quantities of fissile nuclear material (including plutonium) and should be well safeguarded in a commercial facility. Idaho National Laboratory (INL) currently operates a pyroprocessing facility for treatment of spent fuel from Experimental Breeder Reactor-II with two such ER systems. INL implements MC&A via a mass tracking model in combination with periodic sampling of the salt and other materials followed by destructive analysis. This approach is projected to be insufficient to meet international safeguards timeliness requirements. A real time or near real time monitoring method is, thus, direly needed to support commercialization of pyroprocessing. A variety of approaches to achieving real time monitoring for ER salt have been proposed and studied to date—including a potentiometric actinide sensor for concentration measurements, a double bubbler for salt depth and density measurements, and laser induced breakdown spectroscopy (LIBS) for concentration measurements. While each of these methods shows some promise, each also involves substantial technical complexity that may ultimately limit their implementation. Yet another alternative is voltammetry —a very simple method in theory that has previously been tested for this application to a limited extent. The equipment for a voltammetry system consists of off-the-shelf components (three electrodes and a potentiostat), which results in substantial benefits relative to cost and robustness. Based on prior knowledge of electrochemical reduction potentials for each of the species of interest, voltammetry can be used to quantify concentrations of a variety of elemental species—including uranium, plutonium, minor actinides, and rare earths. Various methods have been tested by other researchers to date—including cyclic voltammetry, square wave voltammetry, normal pulse voltammetry, etc. In most cases, it has been observed that there is a very limited concentration range for which the output can be readily correlated with concentration in the salt. Furthermore, testing to date has been limited to simple ternary salts with only a single element being quantified. While incomplete for application to MC&A for pyroprocessing, these results lead us to believe that voltammetry can be optimized based on salt properties and fundamental electrochemical rate processes to yield a highly accurate and robust method. This project is divided into four tasks jointly executed by three university research groups. This includes experimental measurement of key physical data on the systems of interest, development of a predictive voltammetry model, experimental validation of the voltammetry model, and design/verification of an optimized measurement method. This project supports the goals of the US-ROK Joint Fuel Cycle Study in addition to the NA-24 Office of the National Nuclear Security Agency and the International Atomic Energy Agency (IAEA).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Analytical results from salt batch 9 routine DSSHT and SEHT monthly samples

Strip Effluent Hold Tank (SEHT) and Decontaminated Salt Solution Hold Tank (DSSHT) samples from several of the “microbatches” of Integrated Salt Disposition Project (ISDP) Salt Batch (“Macrobatch”) 9 have been analyzed for 238 Pu, 90 Sr, 137 Cs, cations (Inductively Coupled Plasma Emission Spectroscopy - ICPES), and anions (Ion Chromatography Anions - IC-A). The analytical results from the current microbatch samples are similar to those from previous macrobatch samples. The Cs removal continues to be acceptable, with decontamination factors (DF) averaging 25700 (107% RSD). The bulk chemistry of the DSSHT and SEHT samples do not show any signs of unusual behavior, other than lacking the anticipated degree of dilution that is calculated to occur during Modular Caustic-Side Solvent Extraction Unit (MCU) processing.

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Solvent Hold Tank Sample Results for MCU-18-301- 303 (July 2018), MCU-18-357-359-360 (August 2018), and MCU-18-402-410 (September 2018) (Quarterly Report)

A trend summary of three Solvent Hold Tank (SHT) monthly samples; MCU-18-301-302-303, MCU-18- 357-359-360 and MCU-18-402-410 are reported. Most of the conclusions are based on the September SHT sample (MCU-18-402-410). Analyses of the September SHT sample (MCU-18-402-410) indicated that the Modifier (Cs-7SB) and the extractant (MaxCalix) concentrations were below their nominal recommended concentrations (169,000 mg/L and 46,400 mg/L respectively) by 5% and 8% respectively. The suppressor (N,N’,N”–tris(3,7-dimethyloctyl)guanidine or TiDG) concentration has decreased since the April 2018 measurement (Modular Caustic-Side Solvent Extraction Unit or MCU resumed processing radioactive salt solution from May through August 2018) to 241 mg/L, well below the minimum recommended concentration (479 mg/L). The apparent “downwards” trends observed in the TiDG, MaxCalix, and Modifier measurements are consistent with the startup (or Proficiency Runs) of MCU on May 22, 2018; new levels are consistent with the concentrations observed when MCU was fully operational in 2016. This analysis confirms the Isopar™ L addition to the solvent on July 2 and August 9, 2018. This analysis also indicates the solvent did require a trim addition. Based on the current monthly sample, the concentrations of Isopar™ L, MaxCalix, and Modifier are sufficient for continuing operation but are expected to decrease with time. The TiDG concentration was below its recommended level. Periodic characterization and trimming additions to the solvent are recommended. The Semi-Volatile Organic Analysis (SVOA) did not detect any impurities. However, the Fourier Transform Hydrogen Nuclear Magnetic Resonance (FT-HNMR) analysis detected a low concentration of sec-butyl phenol (SBP) at levels of a few ppm. The impurity concentration was highest in the July sample and consistently detected in the August and September samples (but always a few ppm or less). The source of this impurity is due to the degradation of the Modifier. Another impurity observed in the samples was mercury. Based on the September SHT sample, up to 17 ± 3 micrograms of mercury per mL of solvent was detected (the average of the Direct Mercury Analysis (DMA) and X-Ray Fluorescence (XRF) methods). The higher mercury concentration in the solvent (as determined in the last three monthly samples) is possibly due to the higher mercury concentration in Salt Batches 9 and 10 (SB 10 was recently added to Tank 49H in March 2017). The gamma concentration (~2.3E5 dpm/mL) measured in the September SHT samples was consistent with previous values observed when MCU was fully operational (for example, between December 2016 and January 2017) and it was lower than the April SHT measurement. The “dip” in the gamma measurement for the May 2018 SHT sample was due to Isopar™ L addition to MCU during April and May 2018. The laboratory will continue to monitor the quality of the solvent for any new impurities or degradation of the solvent components.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Decontamination Experiments and Simulations Supporting Recovery After Widespread Contamination - 18685

Clean-up after an uncontrolled nuclear material release differs from common decontamination practices in the nuclear industry and military combat scenarios. Uncontrolled releases, particularly acts of terror, require specific protective actions guides (PAGs) to optimize short and long-term repercussions. Ideally, non-destructive decontamination methods would be employed. To help fulfill this need, the IWATERS (Integrated Wash Aid Treatment Emergency Reuse System) was proposed in 2014 for remediation outside the immediate blast area. 'Wash-Aid' is hosed onto affected areas, and the used wash solution is collected and recycled through sand/clay beds. To help characterize the IWATERS technology, the below was completed: - Decontamination experiments with a pressurized wash solution using Cs-137, Sr- 85, and Eu-152 contaminants on porous urban materials. - Sensitivity analyses on recycling bed variables using GoldSim{sup TM} CT. - A proposed framework that expands the recycling bed model to include probabilistic simulation of immediate recovery efforts. Removal efficacy of tap water and 0.5 M KCl were compared for Cs, Sr, and Eu across asphalt, brick, and concrete. The 0.5 M KCl solution significantly outperformed tap water in every case except for Eu on concrete and Cs on brick. Trends in the high-pressure decontamination with salt additive compared to low pressure decontamination show evidence of physical and chemical removal. Variations in cleaning rate yielded no significant trend. Pressure changes (by adjusting nozzle-to-coupon distance) were only significant for the highest pressure. Hypothetical Contamination Scenario: To tie the decontamination experiments and recycling bed model together, a small contamination scenario was created. Imagine the sides of a small concrete building uniformly contaminated with Cs-137. Power washing the building with 0.5 M KCl solution would result in 857 L of contaminated solution generated in 185 minutes. Two sand clay beds were designed to clean the used solution. With a sand mass fraction of 0.9, a 10 ft{sup 2} bed 4 ft tall could recycle the solution in 1.05 hrs. Changing the sand mass fraction to 0.5 yielded a 2.5 ft{sup 2} bed 4 ft tall (∼55-gallon drum) that recycled the solution in 7.43 hrs. This exercise showed how the sand/clay bed design can easily change to match the primary remediation goals. The second sand/clay bed would reduce the waste generation while the first bed reduces the remediation time. Conclusions: Variations on speed and nozzle distance (pressure) yielded insignificant differences in efficacy except for the closest distance. High-pressure washing showed increased decontamination efficacy with KCl additive in the majority of cases. Except for bed composition, key recycling bed parameters can be varied on-the-fly. This allows for optimization of, for example, time and waste generation. The hypothetical contamination scenario showed pairing decontamination experiment results with modeling tools can lead to increased preparedness for a large-scale contamination event. With collaboration and time, an expanded recovery/remediation model could aid in development of effective protective action guides.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Mineral Recovery Enhanced Desalination (MRED) process: An innovative technology for desalinating hard brackish water

Desalination of water containing high concentrations of calcium, dissolved carbon dioxide, sulfate, silica and other sparingly soluble salts is difficult because of the scaling potential. An innovative pre-treatment scheme was investigated at bench scale that selectively removes these constituents and produces a soft water to enable desalination with high feed water recovery. The process first removes dissolved carbon dioxide by air stripping at low pH. Magnesium is removed by precipitation at high pH. Calcium is removed by ion exchange (IX), and sulfate is removed by nanofiltration (NF). Cation regenerant from IX, containing calcium, and concentrate from NF, containing sulfate, is combined to precipitate gypsum. Concentrate from the desalination process consisting of a concentrated NaCl solution is used to regenerate the IX resins. The selective precipitation, IX, and NF processes were tested in laboratory experiments and produced magnesium hydroxide and gypsum at greater than 90% and 95% purity respectively. A process model was developed to calculate process performance, mass and liquid flow rates. Finally, the MRED process offers the following benefits: 1) greater recovery of brackish feedwater by a desalination process, 2) recovery of marketable commodities and 3) reduction in the volume and mass of waste products from the treatment process.

42 ENGINEERING↗

Effects of Salt Concentration, Ionic Strength, and Water Activity on the Growth of a WIPP Archaeal Isolate, Halobacterium sp.

An extremely halophilic archaeon ( Halobacterium sp., WIPP strain), isolated from WIPP halite, was tested for its ability to grow in media containing various combinations of sodium and magnesium chloride that influenced overall ionic strength and water activity. Halobacterium sp. grew well in most tested salt combinations, and its growth appeared to depend more upon the relative contribution of sodium to overall ionic strength than on other parameters. Growth rates were highest when media consisted of at least 2 M NaCl and lowest when NaCl dropped below 20% of the ionic strength (usually less than 0.5 M). High magnesium concentrations ( 1 and 1.5 M MgCh 2 ) did not prevent Halobacterium sp. growth, as long as the sodium contribution was sufficient. Water activity was not a definitive factor in determining growth, because this value did not reach an extreme lower limit in the experimental media.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Understanding and Controlling Cesium Loading on Crystalline Silicotitanate - 19601

Savannah River Remediation (SRR) contracted with Westinghouse Electric Corporation (WEC), Columbia Energy and Environmental Services (CEES), to build and deliver a modular ion exchange process that will be used to remove radioactive cesium from dissolved salt solution contained in Tank 10 at the Savannah River Site (SRS). The resulting decontaminated salt solution will be sent to Tank 11 and then to Tank 50 for final disposition in grout at the Saltstone facility. At the end of processing, the loaded ion exchange (IX) columns are drained, rinsed, and dried. The IX columns are removed from the modular unit and stored at Interim Safe Storage (ISS). This ion exchange process is called the Tank Closure Cesium Removal (TCCR) system. The TCCR system consists of several unit operations and process monitoring capability housed within three modular units. The first modular unit contains the process control computer, the second modular unit contains the main process equipment (two prefilters, four IX columns, and resin trap (post filtration)), and the third modular unit is the ventilation system. The main unit operations of the TCCR system are filtration, ion exchange, and ventilation. The ion exchange media utilized is a form of inorganic Crystalline Silicotitanate (CST), IONSIV{sup TM} R9120-B supplied by UOP Honeywell. CST has a high selectivity for cesium (Cs) and has some affinity for other Group IA alkali metals on the periodic chart (hydrogen, lithium, sodium, potassium, rubidium, and francium). CST has also been shown to have affinity for strontium and actinides. Cesium loading on the CST is dependent on many factors including chemical composition (cations and anions), temperature, and pH of the salt solution. The actual amount of Cs loaded onto an IX column is dependent on the factors above along with the salt solution volume fed to the IX column. Because of CST's high selectivity for Cs, the Cs-137 equilibrium loading can contribute enough decay heat to result in an ion exchange column boiling event under no flow conditions in the IX column. The boiling event under no flow is a safety concern that must be prevented from occurring. To prevent boiling of the column under no flow conditions, the Cs-137 equilibrium loading must be determined and understood. Several methodologies were evaluated for performing the Cs-137 equilibrium loading determination for CST. After review of available research and development (R and D), batch equilibrium contact test data, and Zheng Anthony Miller (ZAM) model predictions for the salt solution chemistry in Tank 10, a methodology for determining the Cs-137 equilibrium loading on CST was selected. The methodology selected involves an in-tank batch equilibrium contact test performed in duplicate that considers additional corrections to account for anticipated operational evolutions. The in-tank batch equilibrium contact test utilizes ∼0.1 grams of CST secured in a slotted sample vial. The slotted sample vial design allows for salt solution to be in contact with the CST. The slotted sample vials containing the CST are submerged in Tank 10 and exposed to the salt solution for ten days. After ten days, the samples are retrieved and sent to the Savannah River National Laboratory (SRNL) for analysis of the Cs-137 equilibrium loading. Additional corrections are then made to the Cs-137 results to account for anticipated operational evolutions. Anticipated operational evolutions include external additions (water/caustic) to Tank 10 that result in a change to its chemical composition, tank temperature effects, departures from the laboratory testing (mixed vs. stagnant tank conditions), and IX column flushes. A description of the methodologies considered, feasibility of methodology deployment, and the down select process (benefits vs. risks) is discussed in this paper. (authors)

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DFLAW Glass and Feed Qualifications to Support WTP Start-Up and Flow-Sheet Development (Final Report, Rev. 0)

In each test, a distinct feed composition was used to determine the effects of feed composition changes on the retention of technetium and other species as detailed in the Test Plan for this work. Key operating parameters such as the glass temperature (nominally 1150°C) were held constant to investigate the effect of each new feed composition on the retention of technetium. The processing rate was maintained at ≈2,250 kg/m 2 /day of glass, which corresponds to the “stretch-goal” rate of 22.5 MT/day specified by ORP for the WTP LAW melter, or ≈2,000 kg/m 2 /day of glass, which corresponds to the enhanced baseline specified by ORP for the WTP LAW melter (as compared to the nominal 1500 kg/m 2 /day). The bubbling rate was adjusted to achieve a near-complete cold-cap (90-100% of melt surface covered with feed). Quantitative measurements of glass production rates, melter operating conditions (temperatures, pressures, power, flows, etc.), gaseous emissions by Fourier Transform Infrared Spectroscopy (FTIR), and particulate emissions using Environmental Protection Agency (EPA) methods were made for each test. Discharged and melt pool glass samples were inspected for formation of any salt phases and analyzed for chemical composition, including technetium content. Samples of the process fluids from each of the off-gas treatment system components were collected and analyzed. Following each test, the off-gas system was partially disassembled and extensively sampled for any residual feed components residing within the off-gas system. Sampling and analysis during and subsequent to each test was intended to be sufficient for supporting a mass balance for all constituents of interest and particularly technetium, rhenium, iodine, sulfur, chlorine, and fluorine. Additional sampling and analysis of evaporator concentrates and condensates was conducted for various organic compounds.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Recovery and Refining of Rare Earth Elements from Lignite Mine Wastes

The University of North Dakota (UND), in collaboration with a comprehensive team of technical, business and host-site partners, built on prior technology development to complete a front-end engineering and design (FEED) and business planning study to recover and refine rare earth elements (REE) and critical minerals (CM) from North Dakota (ND) lignite mine wastes. The end of project goal was to have an investment quality project and a committed team ready to commercialize the proposed technologies in a future construction and operations phase.

01 COAL, LIGNITE, AND PEAT↗

Electrolytic Oxide Reduction of Plutonium Oxide Surrogates

Our research team has investigated electrolytic oxide reduction (EOR) as an improved metal production process versus the currently employed multiple cycle direct oxide reduction (MCDOR) process for the conversion of plutonium oxide to plutonium metal. Inherent advantages in safety, waste generation, efficiency, and operational ease make EOR an attractive alternative to traditional metal production practices. Phase I of the project consisted of the construction of a prototype electrochemical cell, and proof of concept experiments towards the reduction of SnO 2 and CeO 2 compounds. Phase II of the project was focused on the optimization of the cerium system in preparation for future plutonium EOR experiments. We observed the electrolytic reduction of cerium oxide resulting in mixed degrees of success, and discovered this was a result of parasitic side reactions and partial reduction. We employed cyclic voltammetry (CV) to identify the products of these side reactions and characterize partially reduced species. We also observed that, when metal was successfully produced by EOR, a large amount of passed charge was required. We concluded that this observation was due to several inefficiencies of the process relating to the reduction mechanism (e.g. inefficient use of cathodically produced calcium). We used our CV results to guide our optimization of the EOR of CeO 2 by refining our electrochemical method, cell geometry, and materials of construction. Finally, we developed a next generation cell, worked to establish a new molten salt electrochemical capability in the Plutonium Facility, and completed a strategy for plutonium experiments. Our electrochemical method to produce plutonium metal clearly supports the core mission of the laboratory. Furthermore, this project has allowed for the development of capabilities that will provide valuable data to programs interested in characterization and detection, as well as science of signatures. Additionally, establishing this capability in the plutonium facility will facilitate future safeguard studies that can assist with safeguard development in molten salt reprocessing flowsheets, and molten salt reactor designs.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Crack Growth Rate and Large Plate Demonstration of Chloride-induced Stress Corrosion Cracking in Spent Nuclear Fuel Storage Canisters

This report describes the body of work involving testing and analyses performed in FY18 and FY19 at the Savannah River National Laboratory and at Korea University to determine chloride-induced stress corrosion crack (CISCC) growth rates in weldments of austenitic stainless steel plate material. The primary work involved: several separate crack growth rate test campaigns with new testing setups using laboratory-scale corrosion specimens; analyses to estimate the stress intensity factors for flaws postulated to be present in the residual stress fields of axial and circumferential weldments in a prototypic dry storage canister; and establishment of a “large plate” (approximately 51×46 cm with thickness 16 mm and weight about 30 kg) CISCC test that uses a plate with a circumferential weldment harvested from the Sandia National Laboratories’ full-size mockup dry storage canister. The tests for crack growth rate were conducted on ASTM E1861 bolt-load compact tension specimens in a setup designed to allow initially dried salt deposits to deliquesce and infuse the brine to the crack front under conditions relevant to the canister storage environments (e.g., temperature and humidity). The crack growth rate test results matched previous experimental data in the open literature. These literature data had been compiled by SNL and EPRI to develop a temperature-dependent (K-insensitive) crack growth rate model to support flaw disposition in the ASME Boiler and Pressure Vessel Section XI Code Case N-860, “Examination Requirements and Acceptance Standards for Spent Nuclear Fuel Storage and Transportation Containment Systems.” The knowledge of the salt, temperature, humidity conditions for CISCC including the threshold stress intensity factor, $K_{ISCC}$, were used to setup the large plate test in which a set of partthrough- wall and through-wall electrical discharge machined (EDM) defects were positioned in the weld residual stress field of the welded plate material from the mockup canister. A salt solution was applied to this plate specimen with an air-brush spray, then dried, and then exposed to a constant 75% RH condition at an ambient room temperature of 22 °C. The salt loading and environmental condition is expected to be aggressive to cause CISCC. The visual results at the surface of the plate show corrosion product staining (brown rust appearance) but with no CISCC emanating from the machined notches after approximately 5 months exposure to date. The plate test will continue in FY20 and phased-array ultrasonic testing and other NDE techniques will be applied to identify and characterize CISCC that may emanate from the machined defects. The notches and any CISCC from them will be sectioned at the conclusion of the test to verify NDE characterization, and to develop effective through-wall crack growth rates from cracks that would likely be branched.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Analysis of Sludge Batch 10 Samples from Tank 51H (HTF-51-18-94, -95)

Savannah River National Laboratory characterized sludge samples from Tank 51H that is being used to generate Sludge Batch 10. The composite slurry sample from Tank 51H was analyzed and determined to contain a concentrated salt solution with a sodium concentration of 5.79 M (1.33E+05 mg/L) and insoluble sludge solids content of 7.20 wt% (using all nine replicates from the 3 wt% total solids measurements). Hydroxide is the main anionic species in the salt solution with a concentration of 3.00 M followed by nitrate (0.728 M) and nitrite (0.657 M). The sum of the major cations versus the sum of the major anions from the analysis of the filtered supernatant liquid phase shows a difference of <5% indicating good data quality. The results from the analysis of the digestions of the well mixed composite slurry sample show sodium to be the main component of the total dried solids of the sample followed by aluminum, iron, manganese, and thorium. Additional measurements of the weight percent total solids and density using hand shaking of the sample bottle and recirculation through a peristaltic pump indicate both methods provide adequate mixing of the sludge slurry sample.

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DECOVALEX-2023, Task F Specification, Revision 8

This report is the revised (Revision 8) Task F specification for DECOVALEX-2023. Task F is a comparison of the models and methods used in deep geologic repository performance assessment. The task proposes to develop a reference case for a mined repository in a fractured crystalline host rock and a reference case for a mined repository in a salt formation. Teams may choose to participate in the comparison for either or both of the reference cases. For each reference case, a common set of conceptual models and parameters describing features, events, and processes that impact performance will be given, and teams will be responsible for determining how best to implement and couple the models. The comparison will be conducted in stages, beginning with a comparison of key outputs of individual process models, followed by a comparison of a single deterministic simulation of the full reference case, and moving on to uncertainty propagation and uncertainty and sensitivity analysis. This report provides background information, a summary of the proposed reference cases, and a staged plan for the analysis.

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Influence of Nanoconfinement and Elevated Temperatures on Geocolloid-Facilitated Transport of Energy-related Contaminants

The main objective of this project is to obtain a better understanding of the transport behavior and interactions of geocolloids in the presence of energy- related contaminants under bulk and nanoconfined conditions. In that, we first established solid protocols to synthesize and fabricate numerous types of geocolloids with a. various sizes ranging from 40 – 800 nm and b. different hydrophile-lipophile balance (HLB) ratios in the range of 25:75 to 75:25. Direct force measurements with geocolloids having different degrees of surface coverages (i.e. HLB to realistically mimic adsorption of energy related contaminants) were conducted using the Surface Forces Apparatus (SFA) over a distance regime starting from 8μm all the way down to molecular contact. Repulsive forces were observed on approach starting from > 3μm, followed by an exponential increase of which magnitude appears to be larger than a decay length obtained from Derjaguin–Landau–Verwey–Overbeek (DLVO) theory in pure water. When the geocolloids were confined in salted water, the magnitude of onset of repulsion was varied as a function of salinity in solution, which can significantly alter a purely repulsive screened electrostatic (coulombic) interaction arising from, among geocolloids as well as between geocolloids and geosurfaces. The viscosity and flow characteristics of geocolloidal suspensions at different degrees of confinement were also investigated where we identified highly discontinuous rheological behaviors below a critical nanoconfinement level. We anticipate that the knowledge gained through this study will enable the scientists and researchers to better assess transport and fate behaviors of geocolloidal dispersions that can carry energy-related contaminants under realistically emulated geosystem.

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