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At least 541 records · Page 30

DECOVALEX-2023: Task A Final Report

Task A, also known as HGFrac, examines the fracturing processes that may occur in the Callovo-Oxfordian claystone (COx) in the context of the high-level (HLW) and intermediate-level long-lived (ILW-LL) radioactive waste repository in France. Understanding these processes and improving numerical models to reproduce them will aid in the design, optimization, and safety of the repository. Heat and gas fracturing are studied in two independent subtasks following a stepwise approach: laboratory tests/benchmark exercises, in-situ experiment, and finally, an application case. The in-situ heater experiment aimed to thermally induce a hydraulic fracture; temperature and pore pressure were monitored to detect any evidence of fracturing. The in-situ gas injection experiment aimed to study the effect of the stress orientation and gas injection kinetics on the gas fracturing process. The occurrence of fracturing was monitored by gas pressure measured in the injection interval. In both experiments, the excavation-induced fracture network around the heater/injection boreholes played an important role in the reduction of the compressive stress state, leading to both a tensile and shear failure response of the COx. During the first two years of the project, the research teams working on each task developed and/or proposed numerical approaches for reproducing the occurrence of fracturing in the in-situ experiments. The failure criteria were defined by reproducing the measurements from laboratory extension tests for the heat fracturing subtask. In the gas fracturing subtask, their approaches were used for simulating several benchmark exercises, and an inter-comparison between models was carried out. In both tasks, the developed approaches were compared with a simplified approach considering poro-elasticity for the mechanical behaviour of the COx. Most of the developed approaches are based on a continuous medium that takes into account variations in hydraulic properties due to mechanical degradation, such as plastic deformation or damage. Other approaches implicitly modelled weak planes or embedded discontinuities to reproduce fracture propagation. The potential for fracture initiation was also studied through of a discrete approach. In the second half of the project, the research teams mainly focused on interpretative modelling of two in-situ experiments and a blind prediction exercise to test their respective approaches. The models developed by the research teams were also applied at the repository scale to evaluate fracture initiation in a case study under vi unfavourable conditions, particularly in terms of spacing between High-Level Waste cells. The results showed that the poro-elasticity approach could be an efficient tool for understanding the main processes occurring in the COx. One example is the explicit representation of the excavation-induced fracture network around the boreholes, which yielded acceptable results compared to the measurement data. However, advanced approaches were needed to evaluate the potential increase of the excavation-induced fracture network extend and better understand fracture initiation. The stress analyses carried out by the teams revealed that hydraulic boundary conditions had a strong impact on fracture initiation in the heater experiment. Furthermore, in most cases, the results required higher pore pressure increments to reach fracturing than those measured in the experiment. This implies that the measurements may have been biased by the packer’s capacity to fully isolate the piezometric chambers, leading to lower pressures. On the contrary, there was no agreement on the fracturing mode, as some reported either shear or tensile fracturing, while others reported a combination of the two modes. In the case of gas fracturing, the research teams were limited to the comparison of a single point, which complicated their task. Nonetheless, the numerical results were able to reproduce the measurements and capture processes such as longitudinal gas flow through the excavation-induced fracture network, as suggested by some evidence in the observation piezometric chambers. The numerical models also agreed with the measurements in the sense of higher probability of developing along the injection borehole than radially towards the sound rock. The approaches developed by the research teams showed that they are capable of analysing and reproducing fracture initiation in the COx. However, areas of future work should focus on the fracture propagation and fracture aperture, which were out of the scope of this task. To this end, additional data must be gathered for the parameter characterisation and validation of the numerical models. Nonetheless, various approaches showed promising results as they were able to reproduce fracture development under certain conditions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Product Consistency Test Results for the LAW Phase 5 Glasses

This report provides the results of the Vapor Hydration Test solutions and the Product Consistency Test leachates from the Low-Activity Waste Phase 5 glasses, a series of simulated nuclear waste glasses fabricated at the Pacific Northwest National Laboratory. The series included quenched and canister-centerline cooled versions of the glasses. These data will be used in the development of enhanced property/composition models for waste glass vitrification at Hanford. Several of the blanks had detectable amounts of sodium and/or silicon. The measured concentrations (mg/L) of boron, sodium, and silicon for the low-activity test reference material glass included with the Product Consistency Tests were much lower than the expected low-activity test reference material test results. A review of the Product Consistency Test data indicated that there was generally little difference between the normalized values based on targeted or measured glass composition. Heat treatment had various impacts on the normalized concentration values. Most of the glasses had NC B , NC Na , and NC Si values that were greater than the Hanford Tank Waste Treatment and Immobilization Plant low-activity waste constraint of 4 g/L. The release rates for boron, sodium, and silicon were highly correlated for the study glasses.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Technoeconomic Analysis of a Microwave-Assisted Novel Process That Converts Polypropylene Plastic Waste to Propylene

Propylene is an important petrochemical that is used in various industries, including the automobile and polymer sectors. Conventionally, propylene has been produced through fluid catalytic cracking, steam cracking, and propane dehydrogenation. In this project, a novel process to produce propylene from polypropylene plastic waste using a microwave reactor is introduced. Propylene production is simulated in Aspen Plus for both the conventional propane dehydrogenation process and the novel design, with heat integration applied to minimize utility demands and the conventional process serving as the base case for comparison. Finally, a tailored technoeconomic analysis is carried out for both simulated cases to estimate important economic indicators. Analysis of the results shows that the novel microwave-assisted process outperforms the conventional route for propylene production, achieving a single-pass conversion of almost 100%, compared with only 33.8% in the conventional plant. The Net Present Value of the novel plant is $\$1720.3$ MM, which is 3.5 times higher than the conventional process, while the Levelized Cost of Propylene is reduced by 40%. Capital and operating expenditure are also improved in the proposed scheme, with reductions of approximately $\$29$ MM and $\$222$ MM/yr, respectively. In addition to cost savings, this novel process also provides a convenient means to recycle waste polypropylene.

42 ENGINEERING↗

Roadmap to Iodine and Mercury Abatement Materials Selection in Nuclear Waste Processing Off-Gas Streams

This work provides a guide for candidate mercury (Hg) and iodine-129 (129I) abatement material identification, screening, evaluation, technical gap identification, and bench-scale testing prioritization while also conceptualizing a materials deployment roadmap for implementation of new materials in the Hanford Waste Treatment and Immobilization Plant (WTP) Low-Activity Waste (LAW) Facility at Hanford in Richland, Washington, and elsewhere in the U.S. Department of Energy (DOE) complex for similar applications. The study was prompted by the need to replace the Kombisorb BAT-37 due to uncertainties in Hg and 129I capture performance and future availability for use in the WTP LAW Facility secondary off-gas system, specifically the Carbon Adsorber units. However, replacement of this material could also mitigate two other issues with Kombisorb BAT-37 (and its successor BAT II 37): (1) fire safety risk due to exothermic heat generated by adsorption reactions between the carbon material and Hg and (2) the risk of low retention of 129I in glass and subsequent downstream impacts on secondary liquid waste treatment at the Effluent Treatment Facility.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Upgrading biocrude oil into sustainable aviation fuel using zeolite-supported iron-molybdenum carbide nanocatalysts

Food waste is an underdeveloped source for production of sustainable aviation fuel (SAF). Now, there is no certified conversion process of food waste for SAF by American Society for Testing and Materials (ASTM). We report the use of zeolite-supported molybdenum carbide nanocatalysts in upgrading biocrudes, produced from food wastes through HTL, into SAF precursors. Our data show a complete removal of oxygen from the biocrude through hydrodeoxygenation and a higher heating value of 46.5 MJ/kg, which is comparable to that of Jet A (46.1 MJ/kg). The prescreening tests (tier alpha and beta) show the average carbon number of the distillation cut (150° to 230°C) of upgraded fuel is 10.6, close to the value of 11.4 for average conventional jet fuel, and the specifications of properties including surface tension, viscosity, heating value, flash point, and freezing point were found to meet the standards of SAF. The metal carbide nanocatalysts were reusable in upgrading tests, and the activity of deoxygenation was retained.

10 SYNTHETIC FUELS↗

Flash Recycling of Graphite Anodes

The ever-increasing production of commercial lithium-ion batteries (LIBs) will result in a staggering accumulation of waste when they reach their end of life. A closed-loop solution, with effective recycling of spent LIBs, will lessen both the environmental impacts and economic cost of their use. Presently, <5% of spent LIBs are recycled and the regeneration of graphite anodes has, unfortunately, been mostly overlooked despite the considerable cost of battery-grade graphite. Here, in this work, an ultrafast flash recycling method to regenerate the graphite anode is developed and valuable battery metal resources are recovered. Selective Joule heating is applied for only seconds to efficiently decompose the resistive impurities. The generated inorganic salts, including lithium, cobalt, nickel, and manganese, can be easily recollected from the flashed anode waste using diluted acid, specifically 0.1 M HCl. The flash-recycled anode preserves the graphite structure and is coated with a solid-electrolyte-interphase-derived carbon shell, contributing to high initial specific capacity, superior rate performance, and cycling stability, when compared to anode materials recycled using a high-temperature-calcination method. Life-cycle-analysis relative to current graphite production and recycling methods indicate that flash recycling can significantly reduce the total energy consumption and greenhouse gas emission while turning anode recycling into an economically advantageous process.

36 MATERIALS SCIENCE↗

A review of coal heating value correlations with application to coal char, tar, and other fuels

Accurately measuring a fuel’s heating value is one of the first steps in the classification of a new fuel. Heating values are widely used in coal combustion research and are becoming more useful in other fuel types as well. Many different empirical correlations to predict heating values based on primary organic (CHONS) elemental composition are found in the literature, many of which were originally created to predict heating values of parent coal, and some of which have been extended for biomass. However, no correlations exist for heating values of coal chars and tars. Thirteen literature heating value correlations (10 model forms with 13 unique sets of coefficients) were evaluated for their predictive accuracy of a wide variety of fuel types, including 353 coals. This comparative analysis showed that the predictive capabilities of many of these literature models are very similar for coal-based fuels, with good correlation of coal char heating values. Correlation of coal tar heating values was hampered by the lack of heating value data. The choice of which model to use becomes more important for biomass fuels as well as for more unique fuel types like propellants and explosives, municipal solid waste, etc. Finally, the best model forms for each fuel type are suggested based on several statistical measures of fit.

01 COAL, LIGNITE, AND PEAT↗

Superconductivity out of a non-Fermi liquid: Free energy analysis

In this paper, we present an in-depth analysis of the condensation energy E c for a superconductor in a situation when superconductivity emerges out of a non-Fermi liquid due to pairing mediated by a massless boson. This is the case for electronic-mediated pairing near a quantum-critical point in metal, for pairing in SYK-type models, and for phonon-mediated pairing in the properly defined limit, when the dressed Debye frequency vanishes. We consider a subset of these quantum-critical models, in which the pairing in a channel with a proper spatial symmetry is described by an effective 0 + 1 dimensional model with the effective dynamical interaction V(Ω m ) = g¯ γ /|Ω m | γ , where γ is model-specific (the γ model). In previous papers, we argued that the pairing in the γ model is qualitatively different from that in a Fermi liquid, and the gap equation at T = 0 has an infinite number of topologically distinct solutions, Δ n (ω m ), where an integer n, running between 0 and infinity, is the number of zeros of Δ n (ω m ) on the positive Matsubara axis. This gives rise to the set of extrema of E c at E c,n , of which E c,0 is the global minimum. Here, the spectrum E c,n is discrete for a generic γ < 2 but becomes continuous at γ = 2–0. Here, we discuss in more detail the profile of the condensation energy near each E c,n and the transformation from a discrete to a continuous spectrum at γ → 2. We also discuss the free energy and the specific heat of the γ model in the normal state.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DECOVALEX-2023 Task E Specification (Rev. 0)

This report is the Task E specification (Revision 0) for DECOVALEX-2023. Task E is focused on understanding thermal, hydrological, mechanical and chemical (THMC) processes, especially related to predicting brine migration in heated salt. The main test case being used is the ongoing Brine Availability Test in Salt (BATS) heater test located underground at the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico. This report provides short motivational background, a summary of relevant experiments and data, and a step-by-step plan for the analysis by the teams participating in Task E (Rev. 0 includes detailed description of steps 0 and 1). This document will be revised, and more detail will be added to later steps during DECOVALEX-2023.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Cesium Removal from Surrogate Pyroprocessing Salt by Electrodeposition

Active metals in used nuclear fuel dissolve into the salt during pyroprocessing and are not removed by electrorefining or drawdown operations. The buildup of 137 Cs over time increases the heat load and ionizing radiation level of the salt such that it must be replaced frequently, resulting in a significant amount of salt waste. An effective means of managing cesium in the molten salt electrolyte would increase the efficiency of pyroprocessing and decrease the volume of salt waste requiring disposal. A previous report summarized issues that must be addressed when developing a removal strategy and assessed the suitability of existing methods and remaining technological gaps to their application (Rose and Thomas 2023). Cesium is extremely stable in molten salt as a chloride—even more stable than the LiCl-KCl eutectic base salt used for pyroprocessing fuel—which makes removing cesium a challenge. However, sufficiently strong atomic interactions occur between active metal species and liquid metals that make the electrodeposition of active metal fission products into liquid metal electrodes energetically favorable. The feasibility of recovering cesium from LiCl-KCl pyroprocessing salt through electrodeposition into liquid metals is eing assessed by identifying potentially effective liquid metals and performing tests to determine the effectiveness of electrodepositing cesium from a LiCl/KCl salt into these liquid metals. Previous studies investigating the electrodeposition of Sr 2+ , and Ba 2+ into zinc, cadmium, bismuth, lead, tin and antimony have shown that alkali and alkaline earth metals can be electrodeposited at liquid metal cathodes (Kim et al., 2018). The removal of Ba 2+ and Sr 2+ was measured to be more efficient than the removal of monovalent cations due to the greater thermochemical driving force for alloying those elements with the liquid metal (Jang et al. 2022). Because the equilibrium potentials are dependent on the interactions of the active metal in the liquid metal, it is likely that the other alkali metals Li + , and K + , will deposit from LiCl/KCl salt with the Cs + . Therefore, application of this method to recover active metals from pyroprocessing salt will benefit from the use of a liquid metal and set of operating conditions that sufficiently increase the reduction potential of cesium to remove cesium from the waste salt with an acceptable amount of co-deposited lithium and potassium.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Assessment of Active Metal Removal from Pyroprocessing Salt by Electrodeposition

Active metals present in used nuclear fuel dissolve into the salt during pyroprocessing, but are not recovered during electrorefining or drawdown operations due to their chemical stability in the salt. The accumulation of 137 Cs and 90 Sr in process salt over time increases the heat load and ionizing radiation level of the salt such that it must be replaced frequently, resulting in a significant amount of salt waste. An effective means of removing active metals from the molten salt to enable recycle of the electrolyte (e.g., LiCl/KCl) would increase the efficiency of pyroprocessing and decrease the volume of salt waste requiring disposal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

PERFORMANCE ANALYSIS OF AN ENGINEERING SCALE HYDROTHERMAL LIQUEFACTION SYSTEM

This work evaluates the Modular Hydrothermal Liquefaction System (MHTLS), an engineering-scale, integrated continuous HTL plant operated at the Pacific Northwest National Laboratory (PNNL), for converting realistic wet wastes into energy-dense biocrudes. The production campaigns discussed here processed algae, sewage sludges, lignocellulosic blends, Industrial food waste, and engineered food-waste slurries at 350?°C and around 200?bar, with nominal feed rates of ~12?L?h?¹. We report biocrude yields and composition, establish mass and elemental (C, N) balances, and quantify energy performance via heater duties, heat-exchanger behavior, and system-level efficiencies. Biocrudes contained 76–80?wt?% C (dry, ash-free) with HHVs of 38-41?MJ?kg?¹, substantially higher than feed materials HHVs of 16.6–26.1?MJ?kg?¹ and approaching petroleum fuels. Dry, ash-free biocrude yields of 32–53?wt?% corresponded to 43–71?wt?% carbon yields, with 18–40?wt?% of feed carbon routed to the aqueous phase. Thermal efficiencies were 50-65%, and total energy efficiencies, including reactor heat input, were 35-55%. A counter-current tube-in-tube heat exchanger delivered U values of 200–450?W?m?²?K?¹, with fouling-induced declines impacting heat recovery and heater duty. The analysis highlights three priorities for the process intensification of HTL: robust, fouling-resistant heat recovery, hydrodynamically suitable reactor and heat-exchanger designs, simplified and predictable solids management, and biocrude-water separation.

Biocrude production↗

Investigation of Polyurethane and Fire-Retardant Foams as a Radiological Contamination Fixatives - 20202

Fixation of radiological contamination in decommissioning and deactivation can reduce worker risk, and mitigate potentially hazardous conditions; however, nearly every marketed contamination fixative has been found to be flammable, a significant concern in radiological facilities. Coupled with this, industry fixatives are normally used as a thin coating which can present problems when attempting to stabilize irregular void spaces or areas that are difficult to access and ensure full coverage. To this end, Savannah River National Laboratory (SRNL), in conjunction with Florida International University (FIU), have begun investigating the applicability of polyurethane based foaming materials, which will expand to fill a given asymmetric volume and provide fire retardancy to ensure no release of contamination in the event of a facility fire. Commercial foams undergo exothermic reactions resulting in gaseous release that subsequently causes the curing foam material to expand. Heat generation during curing can potentially compromise rubber parts of an enclosure, a significant concern in nuclear facilities. To mitigate large body heat generation, SRNL has also examined multiple layering of the material to ensure compatibility of the curing material with a previously cured body of the same material. Environmental studies were also completed in order to determine the influence on polyurethane based foams curing process. These studies showed that nearly half of the materials tested delaminated from a cured body of the same material, a property not discussed in commercially available documentation on the products. Of note was the compatibility of the foaming material with the substrate, depending on the type material that the foam was cured in/on, significant delamination could occur. Mechanical testing via tensile pull tester, dynamic mechanical analysis (DMA), and thermogravimetric analysis (TGA) was also performed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Advanced Computational Modeling of High-Level Waste Vitrification at the Hanford Site

The U.S. Department of Energy (DOE) has selected vitrification for stabilizing legacy tank waste at the Hanford site, where radioactive waste from plutonium production was historically stored in underground tanks. This waste will be separated into low-activity waste (LAW) and high-level waste (HLW) fractions and processed at the Waste Treatment and Immobilization Plant (WTP). At WTP, glass melters are used for the vitrification of radioactive tank waste, transforming it into a stable borosilicate glass form for safe long-term storage. The melter vessel is constructed from highly durable and heat-resistant materials, where the vitrification process occurs. The main regions that are modeled are the melt pool, plenum, cold cap, riser/discharge chamber, and surrounding structure with insulation layers. Forced convection induced by air bubblers at the base of the melter ensure uniform temperature distribution and provide heat to the cold cap layer. The cold cap is a region of reacting batch feed that floats on top of the molten glass and is where the batch-to-glass reactions occur. Joule heating provided by electrodes mounted along the vertical walls of the melter and immersed directly in the glass, generates the necessary heat for the net endothermic conversion processes that occur in the cold cap. The high temperatures, radioactivity, and opaque nature of the glass prevent direct observation inside the melters. Therefore, computational models are essential for providing insight into factors that affect melter throughput. Thermocouples in the plenum provide operators with plenum temperature measurements. Operational adjustments include bubbling rate, voltage supplied to the electrodes, feed adjustments, and glass removal rate. Different computational fluid dynamics (CFD) models have been developed, each serving a specific purpose. There are CFD models of different scale melters, as well as models that capture the two-phase flow interfaces of rising bubbles in the molten glass or models with a simplified molten glass region so that the surrounding structure and plenum can be feasibly incorporated. Pilot-scale melter models have been developed to serve as validation of the methods employed in the simulation of the full-scale WTP melters. Models incorporating resolved bubbling are used to develop momentum source terms to implement into a single phase, multi-region, steady-state flow model that is being validated by measured process parameters such as glass production rate, voltage, input power, plenum temperatures, etc. The resolved bubbling model uses the multiphase volume of fluid approach to model the system with a high-resolution interface capturing scheme to maintain sharp interfaces between the molten glass and the air phase. The suite of CFD models is continually being improved to incorporate more realistic physics and achieve faster turnaround time. For example, an incremental controller is implemented to automatically adjust electrode voltage within the simulation to a molten glass set point temperature of 1150°C. Newer models feature improved meshes to ensure conformal meshes between regions and eliminate unnecessary mesh refinement in areas that are not of interest (such as boundary layers in offgas ports). Instead of explicitly modeling the structural, refractory, and insulation layers of the melter, a thermal resistance approach is used with published correlations used for boundary conditions. The development of robust and efficient CFD models will be instrumental in enabling the WTP to successfully fulfill its mission of safely stabilizing legacy nuclear waste.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗

Stability of plutonium oxide nanoparticles in the presence of montmorillonite and implications for colloid facilitated transport

The physical and chemical stability of PuO 2 nanoparticles (intrinsic colloids) in groundwater will control their transport and may affect the performance of high-level radioactive waste repositories. In this study, we examined the chemical stability of two types of PuO 2 nanoparticles at both 25 and 80 °C. The “alkaline” PuO 2 nanoparticles were prepared by neutralizing an acidic Pu(IV) solution with dilute NaOH to pH ~9.5. The “acidic” PuO 2 nanoparticles were precipitated from 0.1 M nitric acid by heating a Pu(IV) solution at 60–80 °C for 30 min. The chemical stability of these PuO 2 nanoparticles was tested in the presence of montmorillonite, a common mineral in the environment and potentially relevant backfill material in some nuclear waste repository designs. The “alkaline” PuO 2 nanoparticles were found to be unstable over a timescale of months at both 25 and 80 °C, with elevated temperature enhancing their dissolution rates and sorption to montmorillonite. PuO 2 nanoparticle dissolution rates decreased with increasing Pu concentration, consistent with solution saturation. The “acidic” PuO 2 nanoparticles appeared to remain stable for much longer than the “alkaline” PuO2 nanoparticles. The “alkaline” PuO 2 nanoparticle dissolution rate constants were as high as 10 -11.9±0.4 mol m -2 s -1 and 10 -11.2±0.2 mol m -2 s -1 at 25 and 80 °C, respectively, while the “acidic” PuO 2 nanoparticle dissolution rate at 80 °C was 10 -13.5 mol m -2 s -1 . Based on transmission electron microscopy, the “alkaline” and “acidic” PuO 2 nanoparticles were of similar size (2.5–4.5 and 2–3 nm nanoparticles, respectively). However, the “acidic” PuO 2 nanoparticles formed more ordered nanoparticle aggregates. Our results suggest that the specific conditions experienced during PuO 2 nanoparticle formation could significantly affect the stability of PuO 2 in the presence of competing sorption processes and, in turn, the relative importance of intrinsic versus pseudocolloid transport in the environment.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

White Paper to Justify the Use of Strip Effluent in the Defense Waste Processing Facility During Slurry Mix Evaporator Processing

At the request of Savannah River Remediation (SRR), a white paper was written to assess whether it would be acceptable to process Strip Effluent (SE) in both the Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) in the Defense Waste Processing Facility (DWPF) for the nitric-glycolic acid flowsheet. Savannah River National Laboratory (SRNL) experimental data and DWPF Sludge Batch 9 (SB9) process data was reviewed, looking at batches which included Precipitate Reactor Feed Tank (PRFT) and/or Strip Effluent Feed Tank (SEFT) feeds that led to long processing times. These batches were reviewed looking for processing problems such as melter feed trips, foamovers, heating rod or steam coil fouling, missed Reduction/Oxidation (REDOX) targets, and other process anomalies.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Spent Crystalline Silicotitanate Storage Study—Post AP-105DF Processing

A Tank-Side Cesium Removal (TSCR) system is under development by Washington River Protection Solutions to support initial production of immobilized low-activity waste (LAW) by feeding Hanford tank supernate from tank farms to the Hanford Waste Treatment and Immobilization Plant (WTP) LAW Facility. Tank waste supernate will be filtered to remove suspended solids and then Cs will be removed by processing it through crystalline silicotitanate (CST) ion exchange media manufactured by Honeywell UOP, LLC. The Cs-loaded CST columns will be stored indefinitely, with a goal of eventual CST removal and treatment. Thus, the spent CST needs to be recoverable after undetermined storage time. Previous testing with AP-105 simulant showed that rinsing the CST bed with 3 bed volumes (1.4 apparatus volumes [AVs]) of 0.1 M NaOH resulted in a dried bed that maintained flow characteristics indicative of ease of recovery. This study explored the intermediate conditions (between feed dried in place and the 1.4 AVs of 0.1 M NaOH rinse) to evaluate CST bed properties after: 1) stoppage with feed in place; 2) stoppage after draining feed; 3) stoppage after 0.7 AV of 0.1 M NaOH rinse through column; 4) stoppage after 1.5 AVs of 0.1 M NaOH rinse through column. Post processing, each column was heated at 50 °C for 19 days under pseudo-storage conditions to simulate the expected dried and stored CST bed conditions. Testing was conducted at the small scale (12-mL bed volume); actual, Cs-depleted, AP-105 tank waste was used as the feed. Post-dried CST bed physical properties (angle of repose and penetration depth) were measured to evaluate how CST moved and flowed. All process stop-conditions resulted in a solidified CST bed except for the final condition, 1.5 AVs of 0.1 M NaOH rinse. At this small scale, the three CST beds presented an issue for retrievability after the short storage period (19 days at 50 °C). The latter case confirmed the results from simulant testing. The testing was intended to provide a preliminary assessment of issues that may arise from desiccation of CST during storage with the indicated salt solutions in place. Since these were small-scale tests, the processing system did not scale to full scale conditions exactly; however, the tests did provide insight into the impact on the dried and stored CST bed after stopping processing at an earlier step (upset condition) than normal. These results indicate that if an upset condition occurs at TSCR, a dilute hydroxide rinse should be considered before the CST dries from internal heating.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗