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At least 433 records · Page 24

EVALUATION OF REQUIREMENTS FOR THE DWPF HIGHER CAPACITY CANISTER

The Defense Waste Processing Facility (DWPF) is considering the option to increase canister glass capacity by reducing the wall thickness of the current production canister. This design has been designated as the DWPF Higher Capacity Canister (HCC). A significant decrease in the number of canisters processed during the life of the facility would be achieved if the HCC were implemented leading to a reduced overall reduction in life cycle costs. Prior to implementation of the change, Savannah River National Laboratory (SRNL) was requested to conduct an evaluation of the potential impacts. The specific areas of interest included loading and deformation of the canister during the filling process. Additionally, the effect of the reduced wall thickness on corrosion and material compatibility needed to be addressed. Finally the integrity of the canister during decontamination and other handling steps needed to be determined. The initial request regarding canister fabrication was later addressed in an alternate study. A preliminary review of canister requirements and previous testing was conducted prior to determining the testing approach. Thermal and stress models were developed to predict the forces on the canister during the pouring and cooling process. The thermal model shows the HCC increasing and decreasing in temperature at a slightly faster rate than the original. The HCC is shown to have a 3°F ΔT between the internal and outer surfaces versus a 5°F ΔT for the original design. The stress model indicates strain values ranging from 1.9% to 2.9% for the standard canister and 2.5% to 3.1% for the HCC. These values are dependent on the glass level relative to the thickness transition between the top head and the canister wall. This information, along with field readings, was used to set up environmental test conditions for corrosion studies. Small 304-L canisters were filled with glass and subjected to accelerated environmental testing for 3 months. No evidence of stress corrosion cracking was indicated on either the canisters or U-bend coupons. Calculations and finite element modeling were used to determine forces over a range of handling conditions along with possible forces during decontamination. While expected reductions in some physical characteristics were found in the HCC, none were found to be significant when compared to the required values necessary to perform its intended function. Based on this study and a review of successful testing of thinner canisters at West Valley Demonstration Project (WVDP), the mechanical properties obtained with the thinner wall do not significantly undermine the ability of the canister to perform its intended function.

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Sample Results from the Interim Salt Disposition Program Macrobatch 8 Tank 21H Qualification Samples

Savannah River National Laboratory (SRNL) analyzed samples from Tank 21H in support of qualification of Macrobatch (Salt Batch) 8 for the Interim Salt Disposition Program (ISDP). An Actinide Removal Process (ARP) and several Extraction-Scrub- Strip (ESS) tests were also performed. This document reports characterization data on the samples of Tank 21H as well as simulated performance of ARP and the Modular Caustic Side Solvent Extraction (CSSX) Unit (MCU). No issues with the projected Salt Batch 8 strategy are identified. A demonstration of the monosodium titanate (MST) (0.2 g/L) removal of strontium and actinides provided acceptable average decontamination factors for plutonium of 2.62 (4 hour) and 2.90 (8 hour); and average strontium decontamination factors of 21.7 (4 hour) and 21.3 (8 hour). These values are consistent with results from previous salt batch ARP tests. The two ESS tests also showed acceptable performance with extraction distribution ratios (D(Cs)) values of 52.5 and 50.4 for the Next Generation Solvent (NGS) blend (from MCU) and NGS (lab prepared), respectively. These values are consistent with results from previous salt batch ESS tests. Even though the performance is acceptable, SRNL recommends that a model for predicting extraction behavior for cesium removal for the blended solvent and NGS be developed in order to improve our predictive capabilities for the ESS tests.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Sample results from the Interim Salt Disposition Program Macrobatch 8 Tank 21H qualification samples

Savannah River National Laboratory (SRNL) analyzed samples from Tank 21H in support of qualification of Macrobatch (Salt Batch) 8 for the Interim Salt Disposition Program (ISDP). An Actinide Removal Process (ARP) and several Extraction-Scrub-Strip (ESS) tests were also performed. This document reports characterization data on the samples of Tank 21H as well as simulated performance of ARP and the Modular Caustic Side Solvent Extraction (CSSX) Unit (MCU). No issues with the projected Salt Batch 8 strategy are identified. A demonstration of the monosodium titanate (MST) (0.2 g/L) removal of strontium and actinides provided acceptable average decontamination factors for plutonium of 2.62 (4 hour) and 2.90 (8 hour); and average strontium decontamination factors of 21.7 (4 hour) and 21.3 (8 hour). These values are consistent with results from previous salt batch ARP tests. The two ESS tests also showed acceptable performance with extraction distribution ratios (D (Cs) ) values of 52.5 and 50.4 for the Next Generation Solvent (NGS) blend (from MCU) and NGS (lab prepared), respectively. These values are consistent with results from previous salt batch ESS tests. Even though the performance is acceptable, SRNL recommends that a model for predicting extraction behavior for cesium removal for the blended solvent and NGS be developed in order to improve our predictive capabilities for the ESS tests.

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Sample Results from Routine Salt Batch 7 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”) 7B have been analyzed for 238 Pu, 90 Sr, 137 Cs, Inductively Coupled Plasma Emission Spectroscopy (ICPES), and Ion Chromatography Anions (IC-A). The results from the current microbatch samples are similar to those from earlier samples from this and previous macrobatches. The Actinide Removal Process (ARP) and the Modular Caustic-Side Solvent Extraction Unit (MCU) continue to show more than adequate Pu and Sr removal, and there is a distinct positive trend in Cs removal, due to the use of the Next Generation Solvent (NGS). The Savannah River National Laboratory (SRNL) notes that historically, most measured Concentration Factor (CF) values during salt processing have been in the 12-14 range. However, recent processing gives CF values closer to 11. This observation does not indicate that the solvent performance is suffering, as the Decontamination Factor (DF) has still maintained consistently high values. Nevertheless, SRNL will continue to monitor for indications of process upsets. The bulk chemistry of the DSSHT and SEHT samples do not show any signs of unusual behavior.

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Mercury Phase II Study - Mercury Behavior in Salt Processing Flowsheet

Mercury (Hg) in the Savannah River Site Liquid Waste System (LWS) originated from decades of canyon processing where it was used as a catalyst for dissolving the aluminum cladding of reactor fuel. Approximately 60 metric tons of mercury is currently present throughout the LWS. Mercury has long been a consideration in the LWS, from both hazard and processing perspectives. In February 2015, a Mercury Program Team was established at the request of the Department of Energy to develop a comprehensive action plan for long-term management and removal of mercury. Evaluation was focused in two Phases. Phase I activities assessed the Liquid Waste inventory and chemical processing behavior using a system-by-system review methodology, and determined the speciation of the different mercury forms (Hg+, Hg++, elemental Hg, organomercury, and soluble versus insoluble mercury) within the LWS. Phase II activities are building on the Phase I activities, and results of the LWS flowsheet evaluations will be summarized in three reports: Mercury Behavior in the Salt Processing Flowsheet (i.e. this report); Mercury Behavior in the Defense Waste Processing Facility (DWPF) Flowsheet; and Mercury behavior in the Tank Farm Flowsheet (Evaporator Operations). The evaluation of the mercury behavior in the salt processing flowsheet indicates, inter alia, the following: (1) In the assembled Salt Batches 7, 8 and 9 in Tank 21, the total mercury is mostly soluble with methylmercury (MHg) contributing over 50% of the total mercury. Based on the analyses of samples from 2H Evaporator feed and drop tanks (Tanks 38/43), the source of MHg in Salt Batches 7, 8 and 9 can be attributed to the 2H evaporator concentrate used in assembling the salt batches. The 2H Evaporator is used to evaporate DWPF recycle water. (2) Comparison of data between Tank 21/49, Salt Solution Feed Tank (SSFT), Decontaminated Salt Solution Hold Tank (DSSHT), and Tank 50 samples suggests that the total mercury as well as speciated forms in the assembled salt batches in Tanks 21/49 pass through the Actinide Removal Process (ARP) / Modular Caustic Side Solvent Extraction Unit (MCU) process to Tank 50 with no significant change in the mercury chemistry. (3) In Tank 50, Decontaminated Salt Solution (DSS) from ARP/MCU is the major contributor to the total mercury including MHg. (4) Speciation analyses of TCLP leached solutions of the grout samples prepared from Tank 21, as well as Tank 50 samples, show the majority of the mercury released in the solution is MHg.

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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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Extraction, -scrub, -strip test results from the interim salt disposition program macrobatch 10 tank 21H qualification samples

Savannah River National Laboratory (SRNL) analyzed samples from Tank 21H in support of qualification of Macrobatch (Salt Batch) 10 for the Interim Salt Disposition Program (ISDP). The Salt Batch 10 characterization results were previously reported.ii,iii An Extraction, -Scrub, -Strip (ESS) test was performed to determine cesium distribution ratios (D(Cs)) and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams; this data will be used by Tank Farm Engineering to project a cesium decontamination factor (DF). This test used actual Tank 21H material, and a sample of the NGS Blend solvent currently being used at the Modular Caustic-Side Solvent Extraction Unit (MCU). The ESS test showed acceptable performance with an extraction D(Cs) value of 110. This value is consistent with results from previous salt batch ESS tests using similar solvent formulations. This is better than the predicted value of 39.8 from a recently created D(Cs) model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Evaluation of Canisterized Foams and Evaluation of Radiation Hardened Foams for D&D Activities

The introduction of polyurethane foams has previously been examined elsewhere within the DOE complex with regards to decontamination and decommissioning (D&D) activities, though its use has been prohibited as a result of excessive heat generation and flammability concerns per the safety basis. Should these foams be found compatible with respect to the facility safety basis requirements, D&D work involving large void containing structures such as gloveboxes could be eased through the fixation of residual contamination after decontamination efforts have concluded. To this end, SRNL embarked on a characterization of commercial epoxy foams to identify the characteristics that would be most important to safety basis requirements. Through SRNL’s efforts, the performance of commercial two-part epoxy foams was evaluated for their foaming characteristics, temperature profiles, loading capability with high-Z (high density) additives, and applicability for shielding gamma emission from isotopes including; Am-241, Cs-137, and Co-60. It was found that these foams are capable of encapsulation of a desired volume, though the ideal and experimental expansion coefficients were found to differ. While heat is generated during the reaction, no samples generated heat above 70 °C. Of the down–selected materials, heating was on the order of 40 °C for the flexible foam and 60 °C for the rigid foam. Both were found to return to room temperature after 20 minutes regardless of the volume of foam cast. It was also found that the direct introduction of high-Z additives were capable of attenuating 98% of Am-241 gamma signal, 16% of Cs-137 signal, and 9.5% of Co-60 signal at 1:1 loading capacities of total liquid constituent weight to additive weight. These efforts are currently being reviewed for the ASTM January 2017 subcommittee discussions to address the lack of test methods and standards regarding these materials with respect to D&D environments.

36 MATERIALS SCIENCE↗

Groundwater Modeling and Predictions of C-14 Transport from Pit 38 at TA-54, Area G

This report summarizes the efforts to understand the impacts of transient flow on infiltration rates through the Pits 37 and 38 disposal units at Material Disposal Area G and on C-14 travel times to the regional aquifer. Water was used in the disposal of bulk waste generated by cleanup activities at MDA B, including to wash the metal bins in which the material was transported to Area G, to decontaminate the containers so they can be returned to service, and to decontaminate the trucks used to transport the bins.

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Cheesecloth Alternatives Testing

Heat source plutonium oxide (HS-PuO 2 ) enriched with 65% or greater plutonium-238, is recovered and purified using an aqueous nitric acid process. Historically, cheesecloth has been employed for small spill clean-up and for routine housekeeping. However, the 2016 Department of Energy (DOE) Operating Experience report, entitled Nitrate Waste Evaluations, stated that the mixing of nitrate wastes with organic absorbents, such as cheesecloth, may render transuranic waste unacceptable for disposal at WIPP in the future. This report explores the feasibility of replacement of cheesecloth with an alternative sorbent wipe that has similar absorption and cleaning abilities and improved acid resistance. Currently, cheesecloth is used for three different functions in HS-PuO 2 operations: spill cleanup; general housekeeping operations such as cleaning glovebox floors and windows, balances, furnaces, etc.; and decontamination of iridium cladded heat sources containing HS-PuO 2 . Alternatives used for spill cleanup must be able to tolerate high neutron doses found in the HS-Pu gloveboxes. They must also be able to withstand exposure to nitric acid concentrations up to 15.8 M without decomposing. Wipes used for general housekeeping must also be durable enough to clean rough surfaces and remain intact when used for scrubbing. Wipes used for clad decontamination do not need to absorb large amounts of liquid but they must be able to withstand neutron exposures, temperatures of up to several hundred degrees Celsius, and exposure to acid solutions followed by mild scrubbing. Finally wipes used for any purpose must resist chemical reactions with nitric acid that could cause issues for waste disposal such as energetic exothermic reactions or flammable gas build-up. The wipes selected for this study were ranked based on their usability (radiation tolerance, absorbency, integrity in nitric acid, and durability) as well as their chemical compatibility with nitric acid as demonstrated using several types of chemical analyses. Wipes were then ranked based on the combination of their usability and chemical compatibility scores.

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Extraction, Scrub, and Strip Test Results from The Interim Salt Disposition Program Macrobatch 11 Tank 21H Qualification Samples

Savannah River National Laboratory (SRNL) analyzed samples from Tank 21H in support of qualification of Macrobatch (Salt Batch) 11 for the Interim Salt Disposition Program (ISDP). The Salt Batch 11 characterization results were previously reported. An Extraction, Scrub, and Strip (ESS) test was performed to determine cesium distribution ratios (D(Cs)) and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams; this data will be used by Tank Farm Engineering to project a cesium decontamination factor (DF). This test used actual Tank 21H material, and a sample of the Next Generation Solvent (NGS) Blend solvent currently being used at the Modular Caustic-Side Solvent Extraction Unit (MCU). The ESS test showed acceptable performance with an extraction D(Cs) value of 117. This value is consistent with results from previous salt batch ESS tests using similar solvent formulations. This cesium removal efficiency is better than the predicted value of 50.8 from a recently created D (Cs) model.

Peters, T.↗

Characterizing The Mechanical Properties of Polyurethane Foams

The implementation of polyurethane foams as a permanent fixative for radioactive contamination is currently being examined within the DOE complex for its use in decontamination and decommissioning (D&D) activities. The work involving the implementation of permanent foaming fixatives in a decontamination scenario would include the filling of structures with large voids, such as gloveboxes, pipes, and tanks, where the use of a traditional 2D paint fixative is impractical or unsafe. The mechanical performance of such commercial off the shelf permanent foaming fixatives is uncharacterized with response to harsh environmental conditions often found in nuclear facilities that are undergoing D&D. To this end, SRNL embarked on a mechanical characterization of 6 polyurethane foams, 4 of which were considered flexible and 2 of which were rigid. Two tests were performed on these foams; a tensile tearing test and a compressive test, both of which were strictly guided by ASTM standards. It was found that in both harsh compressive and tensile scenarios, the Hilti brand foam outperformed all other foams, having a peak stress value higher than the closest competitor by a factor of 2.3, and a compressive yield point that was larger than the closest competitor by a factor of 1.4.

36 MATERIALS SCIENCE↗

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.

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Iodine Flowsheet Status Report

A simplified flowsheet was developed to evaluate the impact of varied assumptions of I-129 decontamination factors on the overall I-129 material balance in the Direct Feed Low-Activity Waste (DFLAW) facility. The flowsheet allows inputs for I-129 feed and Decontamination Factors (DFs) for multiple DFLAW unit operations and provides information on I-129 partitioning to output streams. A series of 6 DFs were evaluated, including baseline DFs from a prior Waste Treatment and Immobilization Plant (WTP) study and refined values based on a Savannah River National Laboratory (SRNL) review. The Microsoft Excel flowsheet spreadsheet will be provided to WRPS to allow further evaluation of the impact of DF assumptions on output streams. The spreadsheet is not specific to iodine, the partitioning of any species can be estimated by entering the assumed DFs for each unit operation. Flowsheet outputs matched results from the prior WTP study and show that different assumptions for the DFs can greatly impact iodine partitioning to primary and secondary waste streams. Results from SRNL's baseline DF values suggest less than 50% of I-129 would be retained in glass versus the 96% when using the BNI DF values. In addition, the flowsheet demonstrated that more than 5 DFLAW operational cycles are required to achieve a steady-state of I-129 to the primary waste stream. It is not possible with the data available to determine whether the BNI approach or the approach taken in this study will be a better match for the iodine partitioning once the LAW facility begins operation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Saltstone Waste Characterization Analysis - Salt Waste Processing Facility (SWPF) Waste Stream (Q4 CY2020)

The Saltstone Production and Disposal Facility is designed and permitted by the state of South Carolina Department of Health and Environmental Control (SCDHEC) to treat and dispose of low-level radioactive and hazardous liquid waste (salt solution) remaining from the processing of radioactive material at the Savannah River Site (SRS). Low-level waste (LEW) aqueous streams from the Effluent Treatment Project (ETP) and decontaminated solutions from the Tank Closure Cesium Removal Unit (TCCR) and the Salt Waste Processing Facility (SWPF) are stored in Tank 50 until the LEW can be transferred to the Saltstone Facility for treatment and disposal. In the past, decontaminated solution from the Modular Caustic Side Solvent Extraction Unit (MCU) was stored in Tank 50 until the LEW could be transferred to the Saltstone Facility for treatment and disposal. MCU is currently in a suspended operations state. LEW that meets the Waste Acceptance Criteria (WAC) can be transferred, stored, and treated in the Saltstone Production Facility (SPF) for subsequent disposal as saltstone grout in the Saltstone Disposal Facility (SDF). Sampling will be conducted as new waste streams are identified for treatment and disposal at the Saltstone Industrial Wastewater Treatment Facility (IWTF) and Z-Area Industrial Solid Waste Landfill (ISWLF), Facility ID# 025500-1603, General Condition B.9 or every six years in accordance with South Carolina (SC) Regulation 61-107.19 Parti C, “Solid Waste Management: Solid Waste Landfills and Structural Fill - General Requirements.”

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Conceptual Model Update of Mercury Sources and Flux at Y-12 and Upper East Fork Poplar Creek, Oak Ridge,Tennessee

The interactions between the key processes that control the fate and transport of mercury (Hg) in the sediments, groundwater, and stream water within the Y-12 National Security Complex (Y-12) in Oak Ridge, Tennessee are complex, and many aspects are not well understood. In contaminated environments such as Y-12, conceptual models are frequently developed to aid in visualizing and understanding the dynamic nature of the hydrologic, geochemical, and physical environment. These conceptual models integrate data in an internally consistent manner to understand processes that control the fate and transport of contaminants. Over the past few decades of environmental investigation at Y-12, a number of conceptual models have been developed to identify and define various technical processes at various scales. The main purpose of the activities described in this report is to update the 2011 conceptual model for Upper East Fork Poplar Creek using the most recent Hg concentration and flux data and informed scientific interpretation. Emphasis was placed on improving the previous model by implementing a more detailed spatial and temporal approach to visualize transport pathways in the watershed and trends in flux and concentration over time, and to compare baseflow and stormflow system dynamics. Detailed descriptions of historical and current Hg sources and transport pathways are also provided for the east and west ends of the facility. This conceptual model will allow the US Department of Energy to evaluate past and present remedial activities and provide a strong technical basis for prioritizing and optimizing remedial responses in a cost-effective and efficient manner. Twenty years ago, Hg flux from Outfall (OF) 200 at the headwaters of East Fork Poplar Creek represented approximately 20% of the overall flux leaving Y-12. By the time of the 2011 conceptual model report, that percentage had increased to 70%–80%. Flux estimates compiled for this report using data from 2009 to 2018 show that OF200 accounted for roughly 60% of the flux leaving Y-12. The relative role of the four storm drain conduits to OF200 appeared similar to 2011, with OF163 being the greatest contributor to downstream flux. An increase in Hg concentrations and flux occurred throughout and downstream of west end storm drains in 2011 in response to a storm drain cleanout. Leading up to the 2011 storm drain cleanout, annual baseflow flux was higher at OF200A6 than at Station 17. From 2011 to 2018, the trend reversed with concentrations and flux being higher at Station 17 than at OF200A6. A spike in concentrations occurred in July 2018, potentially due to a one-time influx of Hg that occurred during COLEX decontamination and decommissioning activities at Y-12. This report includes recommendations for improving our conceptual understanding of Hg sources, transport pathways, and flux at Y-12. Important recommendations include further evaluation of transport pathways, concurrent measurement of Hg concentration and flux under baseflow and stormflow conditions at multiple sites, and research to understand the connections among periphyton and methylmercury concentration, bioavailability, and bioaccumulation. Ongoing facility decontamination and decommissioning efforts, site characterization, remedial actions, and research are facilitating the collection of new Hg data at Y-12. The conceptual model update provided herein takes advantage of the extensive recent Hg sampling to provide an updated assessment of Hg mass balance and support a refined understanding of Hg behavior at or near Y-12. This assessment should assist in future environmental management decisions and in mitigating the impacts of Hg on the surrounding environment.

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Reduced Temperature Cesium Removal from AP-101 Using Crystalline Silicotitanate

The Tank Side Cesium Removal (TSCR) system, currently operational under Washington River Protection Solutions LLC (WRPS), sends initial low-activity Hanford waste tank supernate feed 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 the WTP. Crystalline silicotitanate (CST) ion exchange media, manufactured by Honeywell UOP, LLC (product IONSIV™ R9140-B), was selected as the ion exchange media at TSCR. Laboratory-scale ion exchange processing using TSCR prototypic unit operations continues to contribute toward WRPS establishing accurate process flowsheets for the individual feed campaigns planned for TSCR. This report describes the small-scale ion exchange testing with 14.0 L of diluted and filtered supernate from tank 241-AP-101 (AP-101DF) at 16 °C (62 °F) to demonstrate processing at temperature conditions that are more prototypic of what the TSCR system could experience during colder seasons of the year. Since CST Cs capacity increases with decreasing contact temperature, testing at the lower operating temperature will help to predict the maximum 137 Cs loading onto the CST in the TSCR system. One of the waste acceptance criteria (WAC) for the WTP Low-Activity Waste Facility is that the waste must contain less than 3.18×10 -5 Ci 137 Cs per mole of Na. For the AP-101DF tank waste to meet this criterion, only 0.144% of the influent 137 Cs concentration may be delivered to the WTP; this requires a Cs decontamination factor of 694. Testing with AP-101DF matched TSCR prototypic operations where a lead-lag configuration was used until the lag column reached the WAC limit, then a polish column was brought online for continued processing in a lead-lag-polish column configuration. Feed was processed at 1.9 bed volumes (BVs) per hour; the flowrate, in terms of contact time with the CST bed, matched the expected flowrate at TSCR. The Cs-decontaminated product was retained for vitrification testing (to be reported separately). The lead column reached 62% Cs breakthrough after processing ~1400 BVs of feed; the 50% Cs breakthrough occurred at 1250 BVs. Testing compared to previous AP-107 testing at 16 °C showed ~80 BV increases in volume processed to reach the WAC limit for both lead and lag columns. A similar slope in breakthrough curves for both tests indicates similar kinetic behavior, with variations in feed matrices (Na and Cs concentrations) likely responsible for the deviations in reaching the WAC limit. The Cs effluent from the lag column reached the WAC limit after processing 875 BVs. Anticipating this breakthrough point, the polish column was preemptively installed at 770 BVs. Cs breakthrough from the lag column began at 300 BVs, reaching 5.32×10 0 µCi/mL, or 5.6 % Cs breakthrough, after processing all 1400 BVs of feed. Table S.1 and Figure S.1 summarize the observed column performance and relevant Cs loading characteristics.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Ion Exchange Processing of AN-107 Hanford Tank Waste through Crystalline Silicotitanate in a Staged 2- then 3-Column System

The Hanford Site stores an estimated 56 million gallons of mixed radioactive and chemically hazardous waste in large underground tanks. In support of the Direct Feed Low-Activity Waste (DFLAW) Program for expediting Hanford tank waste supernate treatment, laboratory-scale ion exchange processing using prototypic unit operations was conducted on AN-107 tank waste at the Pacific Northwest National Laboratory Radiochemical Processing Laboratory. This report describes the small-scale ion exchange testing with 13.7 L of diluted and filtered supernate from Tank 241-AN-107 (hereafter referred to as AN-107) at 16 °C (62 °F). One of the waste acceptance criteria (WAC) for the Waste Treatment Plant (WTP) Low-Activity Waste Facility is that the waste must contain less than 3.18×10 -5 Ci 137 Cs per mole of Na. For the AN-107 tank waste to meet this criterion, only 0.147% of the influent 137 Cs concentration may be delivered to the WTP; this requires a Cs decontamination factor of 678. Testing with AN-107 matched current Tank Side Cesium Removal (TSCR) facility prototypic operations where a lead-lag configuration was used until the lag column reached the WAC limit, then a polish column was brought online for continued processing in a lead-lag-polish column configuration. Feed was processed at 1.9 bed volumes (BVs) per hour; the flowrate, in terms of contact time with the crystalline silicotitanate (CST) bed, matched the expected flowrate at TSCR. The Cs-decontaminated product was retained for vitrification testing (to be reported separately). The lead column reached 40% Cs breakthrough after processing ~1700 BVs of feed; the 50% Cs breakthrough was extrapolated from the breakthrough data to occur at 1873 BVs. Testing compared to previous AP-101 and AP-107 testing at 16 °C showed ~300 BV increases in volume processed to reach the WAC limit for both lead and lag columns. The increase in capacity was determined to be due to the significantly lower K concentration in the AN-107 compared to the other tank waste matrices. A comparison in breakthrough curves for the three tests indicated slightly slower kinetic behavior in the AN-107, with variations in feed matrices (high organic complexants) likely responsible for the deviation. The Cs effluent from the lag column reached the WAC limit after processing 1097 BVs. Anticipating this breakthrough point, the polish column was preemptively installed around 900 BVs. Cs breakthrough from the lag column began at 500 BVs, reaching 3.06×10 0 µCi/mL, or 2.6 % Cs breakthrough, after processing all 1700 BVs of feed. Table S.1 and Figure S.1 summarize the observed column performance and relevant Cs loading characteristics.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗