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All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Characterization and CST Batch Contact Equilibrium Testing of Modified Tank 9H Process Supernate Samples in Support of TCCR

The Tank Closure Cesium Removal (TCCR) system uses ion exchange columns filled with crystalline silicotitanate (CST) media to process radioactive waste solutions for the removal of ¹³⁷Cs. TCCR currently focuses on dissolving Savannah River Site (SRS) radioactive tank waste (primarily sodium saltcake solids) within Tank 10H followed by at-tank ion exchange column treatment. Two supernate batches from Tank 10H have been processed through the TCCR unit and processing of a third batch is expected soon. After processing of this supernate batch, plans are to replace the CST columns and process dissolved salt solution from Tank 9H through Tank 10H and then through the new CST columns installed in the TCCR unit. The new columns are expected to contain either a media similar to an archived CST batch (IE-911) or the R9120-B CST media used in the current TCCR columns (the two materials are fundamentally the same; just different specifications, product names, and preconditioning steps). Samples of Tank 9H dissolved saltcake were received at the Savannah River National Laboratory (SRNL) and characterized. The Tank 9H supernate contained a high sodium concentration (~9.6 M Na⁺) and will require dilution to near 6 M [Na⁺] prior to processing through the TCCR unit. The cesium concentration in Tank 9H is currently significantly higher than was observed with Tank 10H. Three dilutions of the Tank 9H supernate were conducted to mimic possible dilutions that could be conducted in the tank farm prior to TCCR processing using inhibited water, sodium hydroxide, and sodium nitrate solutions. Dilution #1 was prepared by diluting the Tank 9H supernate by a factor of 1.6 with inhibited water, Dilution #2 was prepared by diluting the Tank 9H supernate by a factor of 3.7 with a mixed sodium hydroxide/sodium nitrate diluent, and Dilution #3 was prepared by also diluting by a factor of 3.7 but with sodium hydroxide only. All three dilutions have similar Na⁺ concentrations (~6 M), but Dilutions #2 and #3 contain significantly less cesium. Major and key minor components of the diluted Tank 9H supernate samples are provided in Table ES-1. Batch contact equilibrium tests were conducted with the Tank 9H dilutions and the two different CST media batches being considered for use in the new TCCR columns. Results are summarized in Table ES-2 and compared to ZAM model predictions. The highest cesium distribution coefficient and percent removal were observed with Dilution #3. IE-911 CST (an archived CST media batch) was more effective at removing cesium than the more recently prepared R9120-B CST media, though both media samples removed >88% of the cesium and the differences may not be statistically different considering the overall uncertainty. Based on the results, maximum cesium loadings were calculated for each CST media type and Tank 9H dilution. In general, maximum cesium loadings from dilutions of this supernate batch are quite high (approaching 0.1 mmol total Cs⁺/g CST for Dilution #1). The highest calculated maximum ¹³⁷Cs loading for the Tank 9H dilutions using the ZAM model with input of the tank compositions and batch contact results was 207 Ci/kg CST (Dilution #1 with IE-911 CST). In all cases, higher maximum cesium loading values were predicted for IE-911 CST versus R9120-B, although the differences varied considerably between the dilutions. The maximum loading value for IE-911 CST with Tank 9H Dilution #1 was only 7% higher than the maximum loading for R9120-B. The maximum loading value for IE-911 CST with Dilution #3 was 24% higher than the maximum loading for R9120-B. Dilution #2 was intermediate between these values. Note that these maximum loading values are the theoretical calculated values, and actual operating conditions will cause differences. A CST binder dilution (correction) factor near 0.7 (relative to pure powder CST) was required for each batch contact test with IE-911 engineered CST using the three Tank 9H dilutions. Correction factors calculated for R9120-B CST ranged from 0.56 to 0.66 for the three dilutions. Following the batch contact tests, the CST samples were isolated from the Tank 9H solution, washed, dried, and digested in acid following established procedures. The analysis results are provided in Table ES-3. Total cesium loading values determined by CST digestion were similar to the calculated loading values based on solution analysis (Table ES-2) for all samples. In addition, the CST was observed to load calcium (R9120-B sample only), iron, strontium, lead, uranium, and plutonium after contact with the waste supernate, as has been observed previously.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Tank 22 Supernate Sample Characterization for Select Radionuclides

The Savannah River National Laboratory (SRNL) was requested by Savannah River Remediation (SRR), through a Technical Assistance Request (G-TAR-H-00007) to provide additional sample characterization on unfiltered Tank 22 supernate identified as sample HTF-22-18-117, which was delivered to SRNL Shielded Cells on December 14, 2018. A summary of the average analytical results for this unfiltered Tank 22 supernate liquid follows. The density of the unfiltered Tank 22 supernate is 1.008 g/mL, 0.2%RSD. The Tank 22 supernate sample was analyzed for a total of 45 radionuclide species, with 15 of these radionuclides showing activity levels above minimum detectable activity of at least one replicate. Cesium-137 showed the highest activity at 2.90E+07 dpm/mL, 2.35 %RSD (1.30E-02 Ci/ L) with U-235 showing the lowest measurable activity at an average of 1.66E-01 dpm/mL, 6.26 %RSD (7.49E-11 Ci/L). The primary beta emitting radionuclides in the unfiltered Tank 22 supernate sample include Sr-90, Y-90 and Cs-137 at average activity concentrations of 1.10E-05, 1.10E-05 and 1.30E-02 Ci/L, respectively. The primary gamma emitting radionuclide is Ba-137 m at an average activity concentration of 1.23E-02 Ci/L (note that the activity concentration for Ba-137 m was calculated as 94.7% the Cs-137 activity concentration).

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

Assessment of the Corrosion Risks to a Double Shell Tank During the Selective Removal of a Cesium-rich Supernate Layer - 20227

The Direct Feed Low Activity Waste (DF-LAW) activity will provide the initial feed to the Waste Treatment and Immobilization Plant at the Hanford Tank Farm facility. Feed preparation for DF-LAW will be carried out in multiple double shell waste tanks (DST) utilizing the Tank Side Cesium Removal (TSCR) process. Tank AP-106 will be an integral part of the TSCR process and hence key to the DFLAW activity. Prior to being utilized for this purpose, the Cs-137 concentration levels in the AP-106 supernate must be reduced significantly. The AP-106 re-purposing process manages normal waste tank addition procedures such that waste acceptance criteria objectives for Cs-137 removal are achieved. As this process was being developed, several concerns regarding tank integrity during and after repurposing of Tank AP-106, and the associated tanks (i.e., AP-102, AN-101, and AP-108), were addressed. The potential corrosion degradation mechanisms identified based on process knowledge included: general and pitting corrosion, caustic stress corrosion cracking, concentration cell corrosion, and microbiologically induced corrosion (MIC). A systematic approach was undertaken to assess the risk of corrosion during the AP-106 re-purposing process. The assessment considered the initial and present condition of the tank, the service history of the tank, the environmental conditions created in each tank during each stage of the process, process timing, and corrosion testing and modeling that approximated tank conditions. Based on this information, recommendations were provided to minimize the exposure of the tank to corrosive conditions. These recommendations included time limits for each stage of the process, staging of corrosion inhibitor solutions should process delays occur, and waste sampling with potential chemistry adjustments at defined steps in the process. Additionally, recommendations for post-process supernate chemistry adjustments and tank inspections were put forth. The greatest risk for this process occurs during the manipulation of large quantities of Columbia River water (i.e., greater than 450,000 gallons) within AP-106, the associated tanks and the supporting infrastructure (i.e., transfer line piping). This risk becomes more pronounced if there is a process upset and the tank walls remain exposed to a corrosive condition longer than intended. As long as the river water is blended into the supernate in a timely fashion, and the waste is adjusted appropriately with corrosion inhibitors, corrosion may be minimized and the final condition of the supernate in AP-106 and the associated tanks is expected to be in compliance with corrosion chemistry requirements. Contingency plans will be made to minimize the impact of a process upset. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Characterization and CST Batch Contact Equilibrium Testing of Aged TCCR Tank 10H Batch 1A and 2 Process Supernate Samples

The Tank Closure Cesium Removal (TCCR) system uses ion exchange columns filled with Crystalline Silicotitanate (CST) media to process radioactive waste solutions for the removal of Cs-137. TCCR currently focuses on dissolving Savannah River Site (SRS) Tank 10H waste (primarily sodium saltcake solids) within the tank followed by at-tank ion exchange column treatment. Previous equilibrium batch contact tests conducted in the SRNL Shielded Cells laboratory and associated analyses indicated that CST cesium removal performance was lower than predicted by ZAM modeling for TCCR Tank 10H Production Supernate Batch 1A. An average cesium distribution coefficient of 1,948 mL/g which corresponds to 94.1% Cs + removal was observed for Batch 1A supernate at 38 °C in January of 2019 using laboratory pretreated IONSIV™ R9120-B a CST from Lot #209900034 (predominant TCCR CST batch) at a phase ratio of 122 mL/g. This cesium loading was approximately 32% lower than model predictions, presumably due to ion exchange competition or fouling associated with alkaline earth or transition metals present in the solutions at low levels. The current testing was designed to: 1) evaluate whether supernate compositional changes during storage would result in improved cesium removal performance with CST, and 2) assist in media selection and handling protocols for future TCCR columns by evaluating several CST manufacturer batches as well as samples of one CST batch which had been pretreated using two different methods (field and laboratory methods).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Uranium Solubility and Supernate Enrichment Testing for Accelerated Basin De-Inventory Discards to Concentration, Storage, and Transfer Facilities

After future Accelerated Basin De-inventory (ABD) enriched uranium discards into sludge batches (SB), portions of supernate decants during SB preparation will be blended into salt batches (StB) and the resulting feeds must meet the nuclear safety requirements for the Salt Waste Processing Facility (SWPF). During previous SB10 sampling and testing, which involved H-Canyon material containing enriched uranium being mixed with Tank 51 sludge shortly after the H-Canyon stream was neutralized, it was identified that the uranium isotopic enrichment in the supernate deviated from the uranium isotopic enrichment in the slurry. The higher uranium isotopic enrichment in the supernate introduced the risk of challenging the feed requirements of SWPF. The ABD material added to SB11 was isotopically diluted with depleted uranium, mitigating any downstream impacts. However, H-Canyon desires to eliminate or minimize future depleted uranium additions in order to meet the mission schedule.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Plutonium Solubility and Supernate Concentration for Neutralized Fast Critical Assembly Discards to Savannah River Site Tank Waste

The Savannah River Site (SRS) plans to dissolve non-irradiated stainless steel (SS)-clad bundles of Fast Critical Assembly (FCA) materials in eighteen batches.1 FCA dissolution is currently underway in the 6.3D dissolver by simultaneous chemical and electrolytic dissolution, which is required to generate the harsh conditions necessary for dissolution of metal-oxide (MOX) and non-aluminum spent nuclear fuels (NASNFs).2 Nitric acid and potassium fluoride are used to promote chemical dissolution.2 Gadolinium will be added during processing as a thermal neutron poison for criticality control. There are no plans for recovering plutonium from this waste stream. After FCA dissolution, the acidic (HNO3/KF) “discards” containing the dissolved metals will be neutralized by addition of 50 wt% sodium hydroxide to a final free hydroxide concentration of 1.2 M.1 Neutralization will precipitate a slurry of insoluble solids, predominantly metal oxides/hydroxides of plutonium, uranium, and SS components. Small fractions of the SS components, Pu, U, and Gd will remain dissolved in the supernate. The neutralized slurry will be composited to existing radioactive waste storage tanks within the SRS Concentration, Storage, and Transfer Facilities (CSTF) containing other similar sludge batch (SB) materials.1 The fate of soluble plutonium and freshly-precipitated, colloidal plutonium from this process are of concern since the total Pu can challenge the waste acceptance criteria (WAC) at the downstream SRS Liquid Waste (LW) facility. Supernate decants including the neutralized FCA discards (nFCAd) within the CSTF will be composited with salt batch (StB) materials and transferred to the SRS Salt Waste Processing Facility (SWPF), where total plutonium is also of concern.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Progress Report on the Investigation of Separable Floating Layers in Hanford Double-Shell Tank Supernates

A review of studies performed the last four years at 222-S Laboratory on separable floating solids including demonstration of the formation of a floating layer in the organic-bound simulated solids and investigation of floating solids found in AN-101 and AW-104 as well as carbonate precipitation floating on the surface of AN-107 grab samples. Demonstrated that semivolatile organic analysis can detect separable organics including tributyl phosphate, degradation products of Quintolubric hydraulic fluid, and hydrocarbon as tentatively identified compounds. In addition, an evaluation of the relationship of small organic acids to the total organic carbon is discussed in selected double-shell tank supernate. A systematic approach is presented to monitor separable organics and insoluble particulates in the double-shell tank waste.

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

Cesium Exchange onto Crystalline Silicotitanate from Blended Hanford Tank Wastes

The Tank Side Cesium Removal (TSCR) system was developed to filter and remove cesium (Cs and 137 Cs) from Hanford tank waste supernate in preparation for vitrification. The Cs removal will be conducted with crystalline silicotitanate (CST) ion exchange media. Under the planned waste-processing strategy, the tank waste supernate will be queued for TSCR processing in tank 241-AP-107 (AP-107). Once AP-107 tank waste volume is sufficiently depleted, the waste supernate from tank 241-AP-105 (AP-105, the holding tank before transfer to AP-107) will be transferred to tank AP-107. Supernate from another tank will be transferred to the holding tank, AP-105, for eventual transfer to tank AP-107. These supernate streams will undergo blending in tanks AP-107 and AP-105; the volume blend ratios will be driven by how much the tank waste supernate volumes are depleted before the next tank waste is added. The consequence of tank waste blending on Cs uptake by CST was of interest and was tested via batch contacts; results are reported herein.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

FY24 Task 5: Leachate Disposition

Directly feeding sludge solids to the high-level waste (HLW) Waste Treatment Plant represents an alternative flowsheet seeking to initiate sludge processing as soon as possible. Key processing functions previously captured during baseline pretreatment operations include leaching and washing prior to solids concentration. These operations should be considered in the potential direct feed flowsheets to maximize waste feed loading, minimize HLW volume, and mitigate corrosion challenges associated with vitrification of high phosphate and fluoride concentrations. Additionally, single-shell tank (SST) retrievals and waste transfers to double-shell tanks (DSTs) in a direct feed flowsheet would likely also benefit from some level of leaching, washing, and solids concentration in order to reduce DST space and mission duration. These operations could occur in a new facility or potentially in available DSTs. If washing and leaching are utilized, an effective disposition pathway for the wash water and leachate solutions are needed. Three target species that benefit significantly from leaching and washing are phosphate, fluoride and aluminum. Phosphate (PO 4 3- ) and fluoride (F - ) can contribute substantially to the amount of carrier fluid needed for dissolution, and the resulting volume of liquid generated. Disposition of this retrieval solution should be evaluated in order to prevent crystallization of these anions throughout system processing. Since there is a high probability that any retrieval solutions will be at or near their PO 4 3- and F - solubility limits, evaporation or blending with a high Na supernate (>3.5 M) is not recommended for the wash water streams without a method to remove precipitants prior to solution disposal. Additionally, aluminum present in the southeast quadrant of the Hanford site represents roughly 60% of the waste solids in the initial processing tanks. These aluminum solids are in the form of gibbsite (Al(OH) 3 ) and can pose significant challenges for processing due to the fast-settling times and high solids loading associated with these materials. Easily remediated by caustic addition to the solids, these wash solutions could be processed through crystalline silicotitanate (CST) ion exchange columns to prepare the supernate solutions for disposition. The current target for feed conditions to the Low Activity Waste (LAW) melter are waste streams that contain nominally 5-6 M Na. Fractions within the tanks contain upwards of 0.2 M phosphate and fluoride in solution at 3.5 M Na. Concentrating these solutions above 5 M Na would result in an exceedance of the solubility limits, and potential for uncontrolled precipitation of the phosphate and fluoride crystal material. The resulting crystalline salt material is typically sodium fluoride phosphate, also referred to as natrophosphate (Na 7 FPO 4 ·19H 2 O). Salt phases are of importance in tank waste due to their chemical reactivity, which can result in precipitation, dissolution, or transformation, impacting any downstream processes (Bolling et al. 2020, Russell, Snow, and Peterson 2010). Salt generation and precipitation could pose challenges by causing system plugging and melter corrosion if left in the supernate stream, or limit sodium molarity of the supernate that would be accepted without incident in waste operations. To understand the impact of this salt generation, the crystallization of natrophosphate in multiple simulant feed matrices was studied to understand the implications of various tank waste supernate chemistries. Three matrices were examined: high PO 4 3- /low F - , low PO 4 3- /high F - , and an average matrix. Subsequent testing was performed with the average matrix with the inclusion of CsNO 3 , and a final run with the average matrix including CsNO 3 and a 137 Cs spike for tracer purposes.

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

Qualification of Processing Batches for Tank Closure Cesium Removal Through In-Tank Batch Contact Testing - 20470

The Tank Closure Cesium Removal (TCCR) process at the Savannah River Site (SRS) is currently processing waste from Tank 10H, generating decontaminated salt solution that is sent to Tank 11H. Transfers out of Tank 11H are then disposed of on site as a grout wasteform in the SRS Saltstone Processing Facility. This is a full-scale demonstration of the use of Crystalline Silicotitanate (CST) for the decontamination of aqueous nuclear waste supernate. CST primarily removes cesium from the high sodium alkaline feed, but also adsorbs strontium, actinides, and some other trace metals. The TCCR unit is an at-tank set of four ion exchange columns, with two typically being operated in series during waste processing. Tank 10H is serving a dual function as both the salt dissolution tank as well as the feed tank for the TCCR system. Prior to operation of TCCR, Tank 10H must undergo dissolution campaigns, dissolving the salt cake to form an aqueous salt solution (supernate). After each dissolution campaign, the supernate created must be qualified prior to processing through the TCCR system. Qualification includes detailed characterization of the supernate as well as in-tank batch contact tests to determine the equilibrium loading of Cs-137 on the CST ion exchange media (IONSIV{sup TM} R9120-Ba). In support of the in-tank batch contact testing, Savannah River National Laboratory (SRNL) developed a CST sample holder, also referred to as a 'tea-bag', to hold a measured amount (∼0.1 grams) of pretreated CST between stainless steel screens. The tea-bag fits within a standard stainless steel sample vial that has been modified to allow the flow of supernate to the CST. This sample holder is then immersed in the waste tank allowing free contact between the CST beads and the surrounding liquid, at an effectively infinite liquid-to-solid phase ratio. A pair of tea-bags are deployed in the tank for a period of 10 days for each batch, after which the CST is recovered, digested, and analyzed to determine the loading of cesium isotopes on the CST. This result then supports the TCCR column thermal loading limits for processing as part of the TCCR safety basis. This paper will discuss the design and testing of the CST sample holder, as well as results from the first few batches of waste qualified and subsequently processed through the TCCR system. In addition, results from analyses of Tank 11H samples, showing decontamination of the product will be included. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Cesium Ion Exchange Testing Using a Three-Column System with Crystalline Silicotitanate and Hanford Tank Waste 241-AP-107

Washington River Protection Solutions (WRPS) is working to support early production of immobilized low-activity waste by feeding Hanford tank supernate from tank farms to the Hanford Waste Treatment and Immobilization Plant (WTP) Low-Activity Waste (LAW) facility. This goal incorporates the design of a Tank-Side Cesium Removal (TSCR) system, which in effect filters tank waste supernate and processes it through a cesium ion exchange medium to remove 137Cs. The 137Cs-depleted product is sent to the WTP for vitrification. Rovira et al. (2018) reported the successful Cs removal from Hanford tank 241-AP-107 waste (hereafter called AP-107) using crystalline silicotitanate (CST) ion exchange media manufactured by Honeywell UOP, LLC (product IONSIVTM R9140-B, Batch 2081000057) in a lead-lag column system. Since that testing, WRPS has changed the column design to a three-column system. WRPS requested a repeated study with AP-107 tank waste using a different batch of IONSIVTM R9140-B, Lot 2002009604, in a three-column format. A small-scale test platform was constructed to demonstrate solids filtration, Cs removal, and LAW vitrification and installed at Pacific Northwest National Laboratory. This report describes the small-scale ion exchange testing system component of the test platform and the processing of approximately 8.55 L of filtered AP-107 supernate using CST, Lot 8056262-999, that passed through a 25-mesh filter. This report also describes the Cs ion exchange batch contact testing to determine the Cs distribution coefficient and the maximum Cs loading capacity in the AP-107 tank waste matrix. Batch contact testing helps to evaluate CST performance on tank waste supernate and is often used prior to processing tank waste through ion exchange columns to assess some performance metrics (e.g., 50% Cs breakthrough). Batch contact tests were performed with the filtered AP-107 tank waste at four Cs concentrations at a phase ratio of 200 (liquid volume to exchanger mass). The distribution coefficient (K d ) at the equilibrium condition of 9.2 µg Cs/mL (AP-107 feed condition) was determined to be 806 mL AP 107/g CST. With a CST bed density of 1.01 g/mL (<25 mesh CST), this K d corresponded to a predicted 50% Cs breakthrough of 814 bed volumes (BVs). The Cs load capacity at the equilibrium feed condition was determined to be 8.3 mg Cs/g dry CST. The three-column format testing was prototypic to the intended TSCR operations, albeit on a small scale with 10-mL CST beds. Flowrate was adjusted to match the CST contact time expected for the full-scale operation, i.e., matched bed volumes per hour (BV/h) flow rate. The feed was processed downflow through the lead column, then through the middle column, and then through the polish column. Loading continued at an average of 1.88 BV/h (1.83 to 2.04 BV/h range) until the entire available AP-107 feed was processed. The Cs-decontaminated product will be provided for vitrification testing. The lead column only reached 45% Cs breakthrough after processing 855 BVs of feed. The 50% Cs breakthrough was extrapolated to occur at ~900 BVs. This extrapolated 50% Cs breakthrough value differed from the batch contact estimate (814 BVs) by 13%. The waste acceptance criteria (WAC) or contract limit for the WTP LAW vitrification facility is 3.18E-5 Ci 137 Cs per mole Na. For the AP-107 tank waste, only 0.127% of the influent 137 Cs concentration may be delivered to the WTP; this required a Cs decontamination factor of 787. The Cs effluent from the middle column reached the WAC after processing 590 BVs. Cs breakthrough from the polish column began at 590 BVs, reaching 6.3E-3 µCi/mL or 4.3E-3% Cs breakthrough after processing 844 BVs of feed.

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

Trace Chemical Detection Using Intercalated MXenes as a Signal Enhancing Substrate in Optical Probes

MXenes are 2D materials composed of layered transition metal nitrides or carbides. These materials are synthesized by HF exfoliation from MAX phases (Ti{sub 3}AlC{sub 2}). The 2D nanomaterial was synthesized by the removal of the 'A' element, resulting in a Mxene product (Ti{sub 3}C{sub 2}). MXenes have the general formula M{sub n+1}X{sub n}T{sub x}, where M is an early transition metal, X is Carbon and/or Nitrogen, and T accounts for surface terminated functional groups such as Fluoride, hydroxyl, and oxygen. These materials have very unique properties, similar to graphene, that allows them to be applied in a variety of trace detection techniques including surface-enhanced Raman spectroscopy (SERS). MXenes have also been demonstrated to selectively uptake uranyl ion, UO{sub 2}{sup 2+}. If this property can be combined with SERS or fluorescence detection, it may be possible to use MXenes as the basis for an alternative method to kinetic phosphorescence analysis (KPA) for trace uranyl measurements. Objectives: To confirm that MXene Nano materials are suitable substrates for SERS and sensor development by enhancing Raman signaling. To determine if certain MXene preparation methods yield materials that are more suitable for trace sensing methods. To determine uranyl uptake properties of these MXene materials and test them for analytical signals. Sample Preparation: Preparation of Ti{sub 3}C{sub 2}MXene (at FSU). MXenes were prepared by etching Al from Ti{sub 3}AlC{sub 2} (MAX phase)material. Two etching techniques yield different MXene products: LiF/HCl: Milder reaction, larger MXene flakes. HF: Harsher reaction, smaller flakes, larger layer separation. Products washed to remove etchant, vacuum filtered, and dried. Dried MXene flakes are air-stable. Film preparation for sensor testing (at SRNL): Suspend powder in diH{sub 2}O, purge with Ar, sonicate for 30 min. Centrifuge and collect supernate with suspended particles. Observed LiF-etched Mxene yielded a higher density of particles and darker collected solution. Drop-cast (4 ml) supernate onto slides and dried with Ar. For Rhodamine B (RhB) testing, drop-cast 4 ml drops onto Mxene spots and dried with Ar. Scanning Electron Microscopy conditions: 10 kV Beam energy, high vacuum; Working distance of 8 mm; beam penetration depth appx. 4 microns, beam spot size appx. 2 nanometers. Results: Detection of aluminum correlates with bright spots on image. Presence of aluminum shows that LiF/HCl etching was less thorough than HF etching. Trace Cl detection in LiF images suggests incomplete rinsing. HF has smaller feature size, more layer structure, and increased homogeneity, consistent with expectations. Macroscopic Raman spectroscopy measurements: 532 nm excitation, ∼50 mW with a ∼100 micron spot size (InPhotonics RPB probe). Kaiser Optical Holospec f/1.8 spectrometer with cooled (-60 deg.C) Andor iDus OE420 CCD. LiF 1x supernate showed good signal for trace measurements of Rhodamine B. HF and 1/4x LiF supernates showed little Mxene or Rhodamine B signal. Low deposition densities led to excess background signal from glass slides. For LiF film, response is linear with Rhodamine B concentration over range tested. Will retest with Raman microscope (∼1 micron spot size) to characterize SERS of more dilute LiF and HF etched Mxenes. Conclusions: The LiF etched material was more suitable for macroscopic SERS measurements because it was more concentrated, resulting in a thicker film than the HF etched Mxene and diluted LiF sample. However, the other materials may give greater SERS enhancements, which we hope to determine from measurements with the Raman microscope. From characterization with SEM we concluded that the HF etched Mxene is more uniform/homogenous and has smaller particle size than the LiF etched Mxene. There is still aluminum present in both samples indicating that etching wasn't complete, but the removal of the aluminum was more efficient in the HF method. Path Forward: Observe SERS with Raman microscopy, to obtain better signals for the more diluted samples and be able to compare enhancement effects for the different MXenes. Characterize uranyl sorption into MXene films and test Raman and fluorescence signals. Revisit the etching conditions to improve removal of aluminum. FSU and SRNL will continue to collaborate to create and characterize different Mxene materials and test their usefulness for sensor applications.

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

Impact of feed variability on cesium removal with multiple actual waste samples from the Hanford site

Here, the Tank Side Cesium Removal (TSCR) system, under development by Washington River Protection Solutions (WRPS), will send initial low-activity Hanford waste tank supernate feeds to the Hanford Waste Treatment and Immobilization Plant (WTP) Low-Activity Waste (LAW) Facility. In addition to entrained solids removal from the supernate, the primary goal of TSCR is to remove cesium-137 ( 137 Cs) by ion exchange, allowing contact handling of the liquid effluent product at WTP. Crystalline silicotitanate (CST) ion exchange media, manufactured by Honeywell UOP (product IONSIV™ R9140-B), was selected as the ion exchange media at TSCR. This lot of material was found to have superior performance in comparison to historic batches of CST. Ion exchange column and batch contact testing with supernate from Hanford tanks AP-105, AP-107 and AW-102 was performed to assess the impact of feed variability on system performance. These tests demonstrated that batch contact measurements provide a reasonable prediction of column capacities with some deviation in performance attributed to column dynamics. In addition, the variability in CST capacity for cesium in the various actual waste samples does not track with historical understanding of the competing cations, suggesting that tank waste samples contain other components that may significantly impact cesium loading.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Impact of Crystalline Silicotitanate Particle Size on Cesium Removal Efficiency

Washington River Protection Solutions is working on a Tank-Side Cesium Removal (TSCR) system to remove Cs from Hanford tank waste supernate. Removal of 137 Cs from tank waste supernate by TSCR results in a low-activity waste (LAW) form that can be fed directly to the Hanford Waste Treatment and Immobilization Plant (WTP) LAW vitrification facility, where the supernate can be vitrified into a glass form. Crystalline silicotitanate (CST) ion exchange media, manufactured by Honeywell UOP, LLC (product IONSIV TM R9140-B), was selected for use as the ion exchange media to remove Cs at TSCR. The particle size used in testing is expected to impact the kinetic performance of the ion exchange media. Therefore, two single-column systems were tested with CST Lot 2002009604. One column contained <25 mesh sieve cut fraction and the other column contained <35 mesh sieve cut fraction. Flowrate through both test systems was adjusted to match the CST contact time expected for the full-scale operation, i.e., matched bed volumes per hour (BV/h) at 1.83 BV/h. The feed was processed downflow through the columns until ~1000 BVs of feed were processed. The waste acceptance criteria (WAC) limit for the WTP LAW vitrification facility is <3.18E-5 Ci 137 Cs per mole Na. For a tank waste containing 156 µCi/mL 137 Cs and 5.6 M Na, up to 0.114% of the influent 137 Cs concentration may be delivered to the WTP (based on 137 Cs content in AP-107 tank waste); this required a Cs decontamination factor of 876. The WAC limit was used as one of the relevant process parameters, along with the 50% breakthrough point, Cs loading per gram of CST, and the mass transfer zone.

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

Summary of Expedited Results from Samples Supporting Tank Closure Cesium Removal (TCCR) Batch 3

Savannah River Remediation (SRR) is currently operating the Tank Closure Cesium Removal (TCCR) process to remove 137 Cs from tank waste supernate using an ion exchange process. As part of that process, Savannah River National Laboratory (SRNL) receives and analyzes samples in support of the qualification of each batch to be processed. SRNL recently received supernate samples retrieved from Tank 10H as well as in-tank batch contact samples for characterization in support of qualifying Batch 3 for processing through the TCCR unit. SRNL received and characterized a set of dip samples collected from Tank 10H (one surface and one variable depth sample). No solids were observed in either sample and therefore the samples were combined, and a suite of analyses were performed on the composite sample. The density of the combined sample was 1.174 g/mL (0.18 %RSD), which was consistent with the measured sodium concentration of 3.54 M (0.08 %RSD). The 137Cs activity in the sample was 4.75E+07 dpm/mL (4.28 %RSD) and the total Cs concentration was calculated to be 1.54 mg/L (1.16E-05 M). The in-tank batch contact samples consist of 0.1 g of crystalline silicotitanate (CST) contained within a teabag device. Duplicate samples were submerged in Tank 10H supernate for a period of about 13 days, after which time they were retrieved and transferred to SRNL for analysis. At SRNL the CST was rinsed to remove excess salt solution and the CST was then air dried before being digested for analysis. Results of the analysis indicated a 137 Cs loading of 3.66E+10 ± 1.40E+09 dpm/g or 16.5 ± 0.63 Ci/kg CST . This value represents a bounding upper limit as it includes the addition of two sigma uncertainty from replicate analysis of the individual teabag samples as well as the addition of the small amount of 137 Cs activity measured in the rinse solutions. The above values are based on the air-dried mass of CST. Correcting to the true dry mass using a F-factor of 0.8191 results in a maximum loading of 20.2 ± 0.77 Ci/kg CST .

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DOE OSTI2020

Characterization of Tank 11H samples from tank closure cesium removal (TCCR) Batch 3 - intermediate and final samples

Savannah River Remediation (SRR) is currently operating the Tank Closure Cesium Removal (TCCR) process to remove 137 Cs from tank waste supernate using an ion exchange process. The TCCR unit processes dissolved salt from Tank 10H through a series of ion exchange columns containing crystalline silicotitanate (CST) and the effluent is then discharged to Tank 11H. Four interim samples pulled from Tank 11H during and just after the completion of processing of Batch 3 through the TCCR process have been analyzed for 137 Cs activity and density. The 137 Cs activity was found to decrease with each subsequent sample, which is consistent with the addition of decontaminated solution to Tank 11H. When compared to the expected composition from mixing the Tank 10H Batch 3 feed with the material already present in Tank 11H, the bulk chemical composition was as expected. A corrosion control sample collected from Tank 11H in June 2020 showed changes in the chemical composition and 137 Cs activity when compared to the composition measured at the end of Batch 2 processing. As there were no additions made to the tank during this period, these changes were attributed to leaching of the solids present in Tank 11H. Revision 1 of this report contains additional analyses of the fourth interim sample including inductively coupled plasma – mass spectrometry (ICP-MS) results and activities of other radionuclides. These results are consistent with the previously reported results, showing leaching from the Tank 11H heel. Notably, the 90 Sr activity was about 2.7x higher than calculated from a mixture of the original Tank 11H supernate and the Tank 10H supernate as treated by TCCR, suggesting additional leaching of Sr from the Tank 11H heel solids. In addition, Revision 1 includes analysis results for the Batch 3 Post-Production surface and variable depth samples collected in October of 2020. The Post-Production surface sample was similar in composition to the fourth interim surface sample, except for the increased Cs-137 and oxalate concentrations that were presumably associated with leaching of the Tank 11H heel solids.

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DOE OSTI2020

Crystalline Silicotitanate Batch Contact Testing with Ba, Ca, Pb, and Sr

Washington River Protection Solutions is working 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. This goal incorporates the design of a Tank Side Cesium Removal system, which filters tank waste supernate to remove suspended solids and then removes Cs by processing it through crystalline silicotitanate (CST) ion exchange media manufactured by Honeywell UOP, LLC. The 137 Cs-depleted product is intended to be sent to the WTP for vitrification. Processing of actual tank waste supernate showed effectively complete uptake of Sr and Ba by CST and significant uptake of Ca. Further, Campbell et al. (2019) analyzed CST post-column testing and found significant (>1E-2 mmoles/g) uptake of Ca and Pb along with some Ba, Cd, Fe, Sr, and U. This led to concern that selected metals, particularly the +2 cations, Ca, Sr, Ba, and Pb may be consuming Cs exchange sites and possibly reducing CST capacity for Cs. Exchange of +2 cations was assumed to be associated with the M(OH) + species for the metal (M) ion in the caustic solution. A series of batch contact testing was conducted to evaluate the exchange behavior of Ba, Ca, Pb, and Sr onto CST. The Cs exchange behavior was also tested as a benchmark for direct comparisons. The CST was provided in the sodium form by Honeywell UOP, as IONSIV TM R9140-B, Lot 2002009604, 18 x 50 mesh. A <30-mesh aliquot was collected to match the sieve fraction expected for use in upcoming small column test configurations. Kinetic exchange rate and isotherms were measured at metal concentrations benchmarked from the AP-107 tank waste feed composition and as limited by the metal solubility in the alkaline solution. Two simplified matrices were tested: 1) 1.0 M NaOH/4.6 M NaNO 3 and 2) 0.1 M NaOH/5.5 M NaNO 3 ; these matrices represented the expected 5.6 M Na concentration of process feed and served to address the hydroxide concentration effect on exchange behavior.

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