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DOE OSTI2012

Operational and Compositional Factors that Affect the Performance Properties of ARP/MCU Saltstone Grout

The Saltstone Production Facility (SPF) receives low level waste (LLW) salt solution from Tank 50H for treatment and disposal. Tank 50H receives transfers from the Effluent Treatment Project (ETP), the H-Canyon General Purpose Evaporator, and the Actinide Removal Process/Modular Caustic Side Solvent Extraction Unit (ARP/MCU) Decontaminated Salt Solution Hold Tank (DSS-HT). At the SPF, the LLW is mixed with premix (a cementitious mixture of portland cement (PC), blast furnace slag (BFS) and Class F fly ash (FA)) in a Readco mixer to produce fresh (uncured) saltstone that is transferred to the Saltstone Disposal Facility (SDF) vaults. The saltstone formulation (mix design) must produce a grout waste form that meets both placement and performance properties. In previous simulated saltstone studies, multiple compositional factors were identified that drive the performance properties of saltstone made from the projected ARP/MCU salt solution. This composition was selected as salt solution simulant since ARP/MCU is the primary influent into Tank 50H. The primary performance property investigated was hydraulic conductivity since it is a variable input property to the saltstone Performance Assessment (PA) transport model. In addition, the porosity, also referred to as void structure, is another variable that impacts the PA response. In addition, Young's modulus and cured density are other performance properties analyzed in this report; however they are indicators of the performance of saltstone and not direct inputs into the PA. The data from previous studies showed that the largest impact on the performance properties of saltstone was due to curing temperature, followed by aluminate concentration in the salt solution, water to premix ratio and premix composition. However, due to the scope of the previous studies, only a few mixes were cured and analyzed at higher temperatures. The samples cured at 60°C had an increased hydraulic conductivity of approximately 600 times that of the sample cured at room temperature. The hydration reactions initiated during the mixing of the premix and salt solution continue during the curing period in the vaults to produce the hardened waste form product. The heat generated from exothermic hydration reactions results in a temperature increase in the vaults that depends on the composition of the decontaminated salt solution being dispositioned, the grout formulation (mix design) and the pour frequency and volume. This heat generation is a contributing factor to the temperature increase in the vaults that leads to an increased cure temperature for the grout. This report will further investigate the impact of curing temperature on saltstone performance properties (hydraulic conductivity, Young's modulus, porosity, etc.) over a range of aluminate concentration, water to premix (w/p) ratio and weight percent fly ash in the premix processed at the SPF. The three curing temperatures selected for this study were chosen to provide data at fixed cure temperatures that represent measured temperatures in the SDF vaults. This does not represent the conditions in the vault where the temperature of the saltstone is continually changing with time. For example, it may take several days for the saltstone to reach 60°C at a given elevation. Previous results demonstrated that the rates at which a selected curing temperature is reached affect the performance properties. The approach taken in this task, a rapid increase to the curing temperature, may be conservative with respect to decreased performance. Nevertheless, the data will provide a basis from which to determine the impact of curing temperature on saltstone performance as a function of key variables. A statistical evaluation of the results for these mixes will be performed to provide the range, and associated uncertainties, of hydraulic conductivity and other properties over this factor space.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2012

Life extension program for the modular caustic side solvent extraction unit at Savannah River Site

Caustic Side Solvent Extraction (CSSX) is currently used at the U.S. Department of Energy (DOE) Savannah River Site (SRS) for removal of cesium from the high-level salt-wastes stored in underground tanks. At SRS, the CSSX process is deployed in the Modular CSSX Unit (MCU). The CSSX technology utilizes a multi-component organic solvent and annular centrifugal contactors to extract cesium from alkaline salt waste. Coalescers and decanters process the Decontaminated Salt Solution (DSS) and Strip Effluent (SE) streams to allow recovery and reuse of the organic solvent and to limit the quantity of solvent transferred to the downstream facilities. MCU is operated in series with the Actinide Removal Process (ARP) which removes strontium and actinides from salt waste utilizing monosodium titanate. ARP and MCU were developed and implemented as interim salt processing until future processing technology, the CSSX-based Salt Waste Processing Facility (SWPF), is operational. SWPF is slated to come on-line in October 2014. The three year design life of the ARP/MCU process, however, was reached in April 2011. Nevertheless, most of the individual process components are capable of operating longer. An evaluation determined ARP/MCU can operate until 2015 before major equipment failure is expected. The three year design life of the ARP/MCU Life Extension (ARP/MCU LE) program will bridge the gap between current ARP/MCU operations and the start of SWPF operation. The ARP/MCU LE program introduces no new technologies. As a portion of this program, a Next Generation Solvent (NGS) and corresponding flowsheet are being developed to provide a major performance enhancement at MCU. This paper discusses all the modifications performed in the facility to support the ARP/MCU Life Extension. It will also discuss the next generation chemistry, including NGS and new stripping chemistry, which will increase cesium removal efficiency in MCU. Possible implementation of the NGS chemistry in MCU accomplishes two objectives. MCU serves as a demonstration facility for improved flowsheet deployment at SWPF; operating with NGS and boric acid validates improved cesium removal performance and increased throughput as well as confirms Defense Waste Processing Facility (DWPF) ability to vitrify waste streams containing boron. NGS implementation at MCU also aids the ARP/MCU LE operation, mitigating the impacts of delays and sustaining operations until other technology is able to come on-line.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Implementation of Increased Salt and Decreased Sludge Volumes in the Defense Waste Processing Facility - 19597

The Savannah River Site (SRS) Liquid Waste System (LWS) safely stores and treats high-level radioactive waste. The LWS consists of 51 waste storage tanks (eight of which are operationally closed and filled with grout), waste evaporators, treatment facilities, and solidification facilities such as the Defense Waste Processing Facility (DWPF) and Saltstone Production Facility (SPF). The disposition of high-level waste (HLW) into a glass waste form poured into canisters at the DWPF began in 1996. DWPF processed a sludge only flowsheet until 2008 when the Actinide Removal Process (ARP) and Modular Caustic Side Solvent Extraction Unit (MCU) began treating the salt waste from the underground liquid storage tanks by striking with monosodium titanate (MST) to remove actinides, when necessary to meet downstream acceptance criteria, and removing the MST and entrained sludge solids (SS) by filtration before using an organic solvent to strip the cesium from the salt, creating two new concentrated HLW streams. This change caused the DWPF to begin processing under both a sludge-only and a coupled operations flowsheet where salt streams would only be added when available. This strategy allows for continuous glass canister production at DWPF. Recently, DWPF completed a 10-month melter replacement in December 2017 and the Concentrate, Storage, and Transfer Facility (CSTF) has completed repairs to the 3H Evaporator system. These facility outages have caused delays in preparing the next batch of sludge feed for DWPF. Concurrent with these facility outages, the Salt Waste Processing Facility (SWPF) has been undergoing system testing and commissioning, working toward startup. SWPF is designed to process salt waste in the same manner as ARP/MCU, but with much higher throughput volumes produced per unit time. In order to process the SWPF salt streams efficiently, DWPF will implement a coupled operation only flowsheet. Based on the facility outages mentioned above, efforts have been made to identify strategies to minimize sludge volumes and maximize salt volumes in the DWPF until the next sludge batch is ready for processing in DWPF. Savannah River Remediation (SRR) recently performed an evaluation that examines the effects of reducing the volume of sludge and maximizing the addition of the cesium-rich concentrated salt solution in each Sludge Receipt and Adjustment Tank (SRAT) batch in order to extend the existing sludge batch as long as possible until the next sludge batch is ready. These effects include a reduction of weight percent solids in the DWPF processing vessels, an increase in cycle time, a new canister heat generation value that remains within regulatory limits, and higher gamma source that remains within bulk facility gamma shielding. Balancing the above parameters resulted in a small reduction of sludge volume per SRAT batch under SWPF operation and maximizing the salt volumes for both ARP/MCU and SWPF operation. This slight reduction in sludge volume and increase in salt volumes allows SRR to mitigate some of the schedule impact incurred by the facility outages. This strategy was evaluated for both ARP/MCU and SWPF operation. This strategy has been successfully implemented under ARP/MCU operation using small step changes in the current process. This paper provides an overview of the evaluation performed for DWPF with ARP/MCU and DWPF with SWPF operations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Deployment of large tank monosodium titanate strike to increase salt processing at the Savannah river site - 15115

The treatment of high level waste (HLW) salt solutions using the Actinide Removal Process (ARP) and Modular Caustic Side Solvent Extraction Unit (MCU) began at Savannah River Site (SRS) in 2008. Since that time, over 6 million gallons of liquid HLW have been treated and the resulting low level waste (LLW) effluent immobilized in cementitious grout at salt-stone. In addition to the continued stabilization of liquid HLW, operating experience from the ARP/MCU facilities has also been valuable in considering the design and construction of the Salt Waste Processing Facility (SWPF) which will utilize similar unit operations on a much larger scale. ARP removes actinides by contacting waste with monosodium titanate (MST) which is a solid adsorbent for soluble strontium and actinides. Once the sorption reaction is complete, the MST solids are separated from salt solution via ultrafiltration through a crossflow filter. The throughput of initial operations of ARP/MCU was limited by the flowrate through the extraction process. However, after the implementation of a new solvent flowsheet, filtration of MST will be the rate limiting step in salt processing. For the entirety of ARP/MCU operation, Savannah River Remediation (SRR) has sought to improve process efficiency and performance through optimization and modification of the original facility design. These efforts have resulted in a best-single-year throughput of 4.9 million liters of decontaminated salt solution. In keeping with these practices, SRR recently initiated a project to improve filtration throughput by relocating the MST strike from small batches immediately upstream of the filtration unit to the underground waste tank used to batch and store salt solution feed for ARP/MCU. This process change will allow the bulk of MST solids to decant in the waste tank prior to filtration which should improve both the nominal operational filter flux as well as the achieved utility of filter operation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2010

Improved Modular CSSX Unit (MCU) Operations

Tiina Laupa and Mark Geeting of SRS: Savannah River Remediation group discuss the use of Actinide Removal Process/Modular Caustic Side Solvent Extraction Unit (ARP/MCU) at the site, how it has performed, improvements made, and future work. One of these improvements, the use of next-generation solvents (NGS), is further discussed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2011

SALTSTONE 1QCY11 TCLP RESULTS

A Saltstone waste form was prepared in the Savannah River National Laboratory (SRNL) from a Tank 50H sample and Z-Area premix material for the first quarter of calendar year 2011 (1QCY11). After the prescribed 28 day cure, samples of the saltstone were collected, and the waste form was shown to meet the South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-79.261.24 and R.61-79.268.48(a) requirements for a nonhazardous waste form with respect to RCRA metals and underlying hazardous constituents. These analyses met all quality assurance specifications of USEPA SW-846. The Saltstone Production Facility (SPF) receives waste from Tank 50H for treatment. In the first quarter of the 2011 calendar year (1QCY11), Tank 50H accepted transfers of approximately 15 kgal from the Effluent Treatment Project (ETP), approximately 15 kgal from Tank 710 - the H-Canyon General Purpose Evaporator, approximately 73 kgal from the H-Canyon Super Kukla campaign, approximately 285 kgal from the Actinide Removal Process/Modular Caustic Side Solvent Extraction Unit (ARP/MCU) Decontaminated Salt Solution Hold Tank (DSS-HT), and approximately 21 kgal from other sources. The Saltstone Grout Sampling plan provides the South Carolina Department of Health and Environmental Control (SCDHEC) with the chemical and physical characterization strategy for the salt solution which is to be disposed of in the Z-Area Solid Waste Landfill (SWLF). During operation, samples were collected from Tank 50H and grout samples prepared to determine the non-hazardous nature of the grout to meet the requirements of the South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-79.261.24(b) and R.61-79.268.48(a). Savannah River National Laboratory (SRNL) was asked to prepare saltstone from samples of Tank 50H obtained January 5, 2011 during 1QCY11 to determine the non-hazardous nature of the grout. The samples were cured and shipped to Babcock & Wilcox Technical Services Group-Radioisotope and Analytical Chemistry Laboratory (B&W TSG-RACL) to perform the Toxic Characteristic Leaching Procedure (TCLP) and subsequent extract analysis on saltstone samples for the analytes required for the quarterly analysis saltstone sample. In addition to the eight toxic metals-arsenic, barium, cadmium, chromium, mercury, lead, selenium and silver-analytes included the underlying hazardous constituents (UHC) antimony, beryllium, nickel, and thallium which could not be eliminated from analysis by process knowledge. B&W TSG-RACL provided subsamples to GEL Laboratories, LLC for analysis for the UHCs benzene, phenols and total and amenable cyanide.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2011

SALTSTONE 2QCY11 TCLP RESULTS

The Saltstone Production Facility (SPF) receives waste from Tank 50H for treatment. In the second quarter of the 2011 calendar year (2QCY11), Tank 50H accepted transfers of approximately 15 kgal from the Effluent Treatment Project (ETP), approximately 2 kgal from Tank 710 - the H-Canyon General Purpose Evaporator, approximately 63 kgal from the HCanyon Super Kukla campaign, approximately 370 kgal from the Actinide Removal Process/Modular Caustic Side Solvent Extraction Unit (ARP/MCU) Decontaminated Salt Solution Hold Tank (DSS-HT), and approximately 10 kgal from other sources. The Saltstone Grout Sampling plan provides the South Carolina Department of Health and Environmental Control (SCDHEC) with the chemical and physical characterization strategy for the salt solution which is to be disposed of in the Z-Area Solid Waste Landfill (SWLF). During operation, samples were collected from Tank 50H and grout samples prepared to determine the non-hazardous nature of the grout to meet the requirements of the South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-79.261.24(b) and R.61-79.268.48(a). Savannah River National Laboratory (SRNL) was asked to prepare saltstone from samples of Tank 50H obtained April 5, 2011 during 2QCY11 to determine the non-hazardous nature of the grout. The samples were cured and shipped to Babcock & Wilcox Technical Services Group-Radioisotope and Analytical Chemistry Laboratory (B&W TSG-RACL) to perform the Toxic Characteristic Leaching Procedure (TCLP) and subsequent extract analysis on saltstone samples for the analytes required for the quarterly analysis saltstone sample. In addition to the eight toxic metals - arsenic, barium, cadmium, chromium, mercury, lead, selenium and silver - analytes included the underlying hazardous constituents (UHC) antimony, beryllium, nickel, and thallium which could not be eliminated from analysis by process knowledge. B&W TSGRACL provided subsamples to GEL Laboratories, LLC for analysis for the UHCs benzene, phenols and total and amenable cyanide.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2011

RESULTS FOR THE SECOND QUARTER 2011 TANK 50 WAC SLURRY SAMPLE: CHEMICAL AND RADIONUCLIDE CONTAMINANT RESULTS

The Saltstone Facility is designed and permitted to immobilize and dispose of low-level radioactive and hazardous liquid waste (salt solution) remaining from the processing of radioactive material at the Savannah River Site. Low-level waste (LLW) streams from the Effluent Treatment Project (ETP), H-Canyon, the DDA (Deliquification, Dissolution, and Adjustment) process, and the decontaminated salt solution product from the Actinide Removal Process/Modular Caustic Side Solvent Extraction (CSSX) Unit (ARP/MCU) process are stored in Tank 50 until the LLW can be transferred to the Saltstone Facility for treatment and disposal. The LLW must meet the specified waste acceptance criteria (WAC) before it is processed into saltstone. The specific chemical and radionuclide contaminants and their respective WAC limits are listed in the current Saltstone WAC. Waste Solidification Engineering (WSE) requested that Savannah River National Laboratory (SRNL) perform quarterly analysis on saltstone samples. The concentrations of chemical and radionuclide contaminants are measured to ensure the saltstone produced during each quarter is in compliance with the current WAC. This report documents the concentrations of chemical and radionuclide contaminants for the 2011 Second Quarter samples collected from Tank 50 on April 4, 2011 and discusses those results in further detail than the previously issued results report. The following conclusions are drawn from the analytical results provided in this report: (1) The concentrations of the reported chemical and radioactive contaminants were less than their respective WAC targets or limits unless noted in this section. (2) The reported detection limit for {sup 59}Ni is above the requested limit from Reference 2 but below the established limit in Reference 3. (3) The reported detection limit for {sup 94}Nb is above the requested limit from Reference 2; however, it is below the established limits in Reference 3. (4) The reported concentration of {sup 242m}Am is above the target in Listed in Attachment 8.4 of the Saltstone WAC. (5) {sup 247}Cm and {sup 249}Cf are above the requested limits from Reference 2. However, they are below the limits established in Reference 3. (6) The reported detection limit for Norpar 13{sup 5} is greater than the limit from Table 4 and Attachment 8.2 of the WAC. (7) The reported detection limit for Isopar L is greater than the limit from Table 3 of the WAC. (8) Isopar L and Norpar 13 have limited solubility in aqueous solutions making it difficult to obtain consistent and reliable sub-samples. The values reported in this memo are the concentrations in the sub-sample as detected by the instrument; however, the results may not accurately represent the concentrations of the analytes in Tank 50.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2011

RESULTS FOR THE THIRD QUARTER 2011 TANK 50 WAC SLURRY SAMPLE: CHEMICAL AND RADIONUCLIDE CONTAMINANT RESULTS

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2012

Saltstone 3QCY11 TCLP Results

A Saltstone waste form was prepared in the Savannah River National Laboratory (SRNL) from a Tank 50H sample and Z-Area premix material for the third quarter of calendar year 2011 (3QCY11). After the prescribed 32 day cure, samples of the saltstone were collected, and the waste form was shown to meet the South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-79.261.24 and R.61-79.268.48(a) requirements for a nonhazardous waste form with respect to RCRA metals and underlying hazardous constituents. These analyses met all quality assurance specifications of USEPA SW-846. The Saltstone Production Facility (SPF) receives waste from Tank 50H for treatment. In the third quarter of the 2011 calendar year (3QCY11), Tank 50H accepted transfers of approximately 20 kgal from the Effluent Treatment Project (ETP), approximately 236 kgal from the Actinide Removal Process/Modular Caustic Side Solvent Extraction Unit (ARP/MCU) Decontaminated Salt Solution Hold Tank (DSS-HT), and approximately 25 kgal from other sources. The Saltstone Grout Sampling plan provides the South Carolina Department of Health and Environmental Control (SCDHEC) with the chemical and physical characterization strategy for the salt solution which is to be disposed of in the Z-Area Solid Waste Landfill (SWLF). During operation, samples were collected from Tank 50H and grout samples prepared to determine the non-hazardous nature of the grout to meet the requirements of the South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-79.261.24(b) and R.61-79.268.48(a). Savannah River National Laboratory (SRNL) was asked to prepare saltstone from samples of Tank 50H obtained July 7, 2011 during 3QCY11 to determine the non-hazardous nature of the grout. The samples were cured and shipped to Babcock & Wilcox Technical Services Group-Radioisotope and Analytical Chemistry Laboratory (B&W TSG-RACL) to perform the Toxic Characteristic Leaching Procedure (TCLP) and subsequent extract analysis on saltstone samples for the analytes required for the quarterly analysis saltstone sample. In addition to the eight toxic metals - arsenic, barium, cadmium, chromium, mercury, lead, selenium and silver - analytes included the underlying hazardous constituents (UHC) antimony, beryllium, nickel, and thallium which could not be eliminated from analysis by process knowledge. B&W TSG-RACL provided subsamples to GEL Laboratories, LLC for analysis for the UHCs benzene, phenols and total and amenable cyanide.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2012

Results for the Fourth Quarter 2011 Tank 50 WAC Slurry Sample: Chemical and Radionuclide Contaminant Results

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2012

Saltstone 4QCY11 TCLP Results

The Saltstone Production Facility (SPF) receives waste from Tank 50H for treatment. In the fourth quarter of the 2011 calendar year (4QCY11), Tank 50H accepted transfers of approximately 10 kgal from the Effluent Treatment Project (ETP), approximately 4 kgal from 211H, approximately 573 kgal from the Actinide Removal Process/Modular Caustic Side Solvent Extraction Unit (ARP/MCU) Decontaminated Salt Solution Hold Tank (DSS-HT), and approximately 5 kgal from other sources. The Saltstone Grout Sampling plan provides the South Carolina Department of Health and Environmental Control (SCDHEC) with the chemical and physical characterization strategy for the salt solution which is to be disposed of in the Z-Area Solid Waste Landfill (SWLF). During operation, samples were collected from Tank 50H and grout samples prepared to determine the non-hazardous nature of the grout to meet the requirements of the South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-79.261.24(b) and R.61-79.268.48(a). Savannah River National Laboratory (SRNL) was asked to prepare saltstone from samples of Tank 50H obtained Oct. 12, 2011 during 4QCY11 to determine the non-hazardous nature of the grout. The samples were cured and shipped to Babcock & Wilcox Technical Services Group-Radioisotope and Analytical Chemistry Laboratory (B&W TSG-RACL) to perform the Toxic Characteristic Leaching Procedure (TCLP) and subsequent extract analysis on saltstone samples for the analytes required for the quarterly analysis saltstone sample. In addition to the eight toxic metals - arsenic, barium, cadmium, chromium, mercury, lead, selenium and silver - analytes included the underlying hazardous constituents (UHC) antimony, beryllium, nickel, and thallium which could not be eliminated from analysis by process knowledge. B&W TSG-RACL provided subsamples to GEL Laboratories, LLC for analysis for the UHCs benzene, phenols and total and amenable cyanide.

36 MATERIALS SCIENCE↗

DOE OSTI2012

Determination of the impact of glycolate on arp and mcu operations

Savannah River Remediation (SRR) is evaluating an alternate flowsheet for the Defense Waste Processing Facility (DWPF) using glycolic acid as a reductant. An important aspect of the development of the glycolic acid flowsheet is determining if glycolate has any detrimental downstream impacts. Testing was performed to determine if there is any impact to the strontium and actinide sorption by monosodium titanate (MST) and modified monosodium titanate (mMST) or if there is an impact to the cesium removal at the Modular Caustic-Side Solvent Extraction Processing Unit (MCU). Sorption testing was performed using both MST and modified MST (mMST) in the presence of 5,000 and 10,000 ppm (mass basis) glycolate. 10,000 ppm is the estimated bounding concentration expected in the DWPF recycle stream based on DWPF melter flammable gas model results. The presence of glycolate was found to slow the removal of Sr and Pu by MST, while increasing the removal rate of Np. Results indicate that the impact is a kinetic effect, and the overall capacity of the material is not affected. There was no measurable effect on U removal at either glycolate concentration. The slower removal rates for Sr and Pu at 5,000 and 10,000 ppm glycolate could result in lower DF values for these sorbates in ARP based on the current (12 hours) and proposed (8 hours) contact times. For the highest glycolate concentration used in this study, the percentage of Sr removed at 6 hours of contact decreased by 1% and the percentage of Pu removed decreased by nearly 7%. The impact may prove insignificant if the concentration of glycolate that is returned to the tank farm is well below the concentrations tested in this study. The presence of glycolate also decreased the removal rates for all three sorbates (Sr, Pu, and Np) by mMST. Similarly to MST, the results for mMST indicate that the impact is a kinetic effect, and the overall capacity of the material is not affected. The presence of glycolate did not change the lack of affinity of mMST for U. Pre-contacting the MST or mMST with glycolate did not have a significant effect on the performance of the materials when compared to tests having the same concentration of glycolate present in the simulant. These findings suggest that the glycolate is likely influencing removal by sorbate complexation and not by depositing onto or forming a film on the surface of the MST solids. Since the DF values are salt batch dependent, it is not possible to a priori quantify the impacts of glycolate on future processing campaigns. However, we recommend that the impacts of glycolate be evaluated during each salt batch qualification when a final processing concentration is defined, and recommendations can then be made on how to mitigate negative impacts, if needed. Impacts to the performance of the MST or mMST could be mitigated by increasing contact time or increasing sorbent concentrations. In addition to the MST and mMST testing, testing was performed to determine if there is an impact to the cesium removal at Modular Caustic-Side Solvent Extraction Processing Unit (MCU). An Extraction-Scrub-Strip (ESS) test routine was used to simulate cesium removal at the MCU. For this, SRNL performed three ESS tests, using the same basic aqueous waste simulant and solvent. For one test, SRNL added 5,000 ppm (mass basis) of glycolate and added 10,000 ppm of glycolate to a second test. A control test contained no glycolate. The results of all three tests were virtually identical for all the extraction, scrub and strip tests. (A single data point in the 5,000 ppm test is physically impossible and SRNL is currently resolving this obvious error.) At this time, SRNL concludes that the presence of up to 10,000 ppm of glycolate does not affect cesium removal by the current solvent system used in the MCU. Although not tested, the impact of glycolate for the Next Generation Solvent - that replaces BOBCalixC6 with MaxCalix - is expected to be very similar to that for the baseline solvent. Testing is needed to confirm. Additional testing is recommended to both further examine the nature of the interaction of glycolate with MST and mMST and also to help address some postulated risks on ARP/MCU operations with implementation of the glycolate flowsheet. The additional testing includes FTIR and Raman spectroscopy to examine the surface of the MST and mMST particles, iodometric titrations to determine the peroxide content in the mMST before and after exposure to glycolate, particle size measurements of the MST and mMST from the experiments performed with glycolate, and measurements of soluble Ti in the supernate from these glycolate experiments. With regards to MCU impacts, the following tests are also recommended based on the premortem risk assessment.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2012

Leaching of Titanium from Monosodium Titanate (MST) and Modified MST (mMST)

Analysis of a fouled coalescer and pre-filters from Actinide Removal Process/Modular Caustic Side Solvent Extraction Unit (ARP/MCU) operations showed evidence of Ti containing solids. Based on these results a series of tests were planned to examine the extent of Ti leaching from monosodium titanate (MST) and modified monosodium titanate (mMST) in various solutions. The solutions tested included a series of salt solutions with varying free hydroxide concentrations, two sodium hydroxide concentrations, 9 wt % and 15 wt %, nitric and oxalic acid solutions. Overall, the amount of Ti leached from the MST and mMST was much greater in the acid solutions compared to the sodium hydroxide or salt solutions, which is consistent with the expected trend. The leaching data also showed that increasing hydroxide concentration, whether pure NaOH solution used for filter cleaning in ARP or the waste salt solution, increased the amount of Ti leached from both the MST and mMST. For the respective nominal contact times with the MST solids - for filter cleaning or the normal filter operation, the dissolved Ti concentrations are comparable suggesting either cause may contribute to the increased Ti fouling on the MCU coalescers. Tests showed that Ti containing solids could be precipitated from solution after the addition of scrub acid and a decrease in temperature similar to expected in MCU operations. FTIR analysis of these solids showed some similarity to the solids observed on the fouled coalescer and pre-filters. Although only a cursory study, this information suggests that the practice of increasing free hydroxide in feed solutions to MCU as a mitigation to aluminosilicate formation may be offset by the impact of formation of Ti solids in the overall process. Additional consideration of this finding from MCU and SWPF operation is warranted.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Results Of Initial Analyses Of The Macrobatch 7 Tank 21H Qualification Samples

Savannah River National Laboratory (SRNL) analyzed samples from Tank 21H in support of qualification of Salt (Macro) Batch 7 for the Interim Salt Disposition Program (ISDP) through ARP/MCU. This document reports the initial results of the analyses of samples of Tank 21H. Further results on the chemistry and other tests will be issued in the future. No issues with the projected Salt Batch 7 strategy are identified, other than the presence of visible quantities of dark colored solids. Based upon a SRNL settling test, the solids should settle well within the months-long settling period to be employed in Tank 21H. However, SRNL recommends analyzing the solids to provide input to OLI modeling in order to evaluate the impacts of these solids to present and future salt batches.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Sample Results From The Interim Salt Disposition Program Macrobatch 7 Tank 21H Qualification Samples

Savannah River National Laboratory (SRNL) analyzed samples from Tank 21H in support of qualification of Macrobatch (Salt Batch) 7 for the Interim Salt Disposition Program (ISDP). An ARP and several ESS tests were also performed. This document reports characterization data on the samples of Tank 21H as well as simulated performance of ARP/MCU. No issues with the projected Salt Batch 7 strategy are identified, other than the presence of visible quantities of dark colored solids. A demonstration of the monosodium titanate (0.2 g/L) removal of strontium and actinides provided acceptable 4 hour average decontamination factors for Pu and Sr of 3.22 and 18.4, respectively. The Four ESS tests also showed acceptable behavior with distribution ratios (D(Cs)) values of 15.96, 57.1, 58.6, and 65.6 for the MCU, cold blend, hot blend, and Next Generation Solvent (NGS), respectively. The predicted value for the MCU solvent was 13.2. Currently, there are no models that would allow a prediction of extraction behavior for the other three solvents. SRNL recommends that a model for predicting extraction behavior for cesium removal for the blended solvent and NGS be developed. While no outstanding issues were noted, the presence of solids in the samples should be investigated in future work. It is possible that the solids may represent a potential reservoir of material (such as potassium) that could have an impact on MCU performance if they were to dissolve back into the feed solution. This salt batch is intended to be the first batch to be processed through MCU entirely using the new NGS-MCU solvent.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

MST Filterability Tests

The Savannah River Site (SRS) is currently treating radioactive liquid waste with the Actinide Removal Process (ARP) and the Modular Caustic Side Solvent Extraction Unit (MCU). The low filter flux through the ARP has limited the rate at which radioactive liquid waste can be treated. Recent filter flux has averaged approximately 5 gallons per minute (gpm). Salt Batch 6 has had a lower processing rate and required frequent filter cleaning. Savannah River Remediation (SRR) has a desire to understand the causes of the low filter flux and to increase ARP/MCU throughput. In addition, at the time the testing started, SRR was assessing the impact of replacing the 0.1 micron filter with a 0.5 micron filter. This report describes testing of MST filterability to investigate the impact of filter pore size and MST particle size on filter flux and testing of filter enhancers to attempt to increase filter flux. The authors constructed a laboratory-scale crossflow filter apparatus with two crossflow filters operating in parallel. One filter was a 0.1 micron Mott sintered SS filter and the other was a 0.5 micron Mott sintered SS filter. The authors also constructed a dead-end filtration apparatus to conduct screening tests with potential filter aids and body feeds, referred to as filter enhancers. The original baseline for ARP was 5.6 M sodium salt solution with a free hydroxide concentration of approximately 1.7 M.3 ARP has been operating with a sodium concentration of approximately 6.4 M and a free hydroxide concentration of approximately 2.5 M. SRNL conducted tests varying the concentration of sodium and free hydroxide to determine whether those changes had a significant effect on filter flux. The feed slurries for the MST filterability tests were composed of simple salts (NaOH, NaNO 2 , and NaNO 3 ) and MST (0.2 – 4.8 g/L). The feed slurry for the filter enhancer tests contained simulated salt batch 6 supernate, MST, and filter enhancers.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Impact Of Sodium Oxalate, Sodium Aluminosilicate, and Gibbsite/Boehmite on ARP Filter Performance

The Savannah River Site (SRS) is currently treating radioactive liquid waste with the Actinide Removal Process (ARP) and the Modular Caustic Side Solvent Extraction Unit (MCU). Recently, the low filter flux through the ARP of approximately 5 gallons per minute has limited the rate at which radioactive liquid waste can be treated. Salt Batch 6 had a lower processing rate and required frequent filter cleaning. Savannah River Remediation (SRR) has a desire to understand the causes of the low filter flux and to increase ARP/MCU throughput. SRR requested SRNL to conduct bench-scale filter tests to evaluate whether sodium oxalate, sodium aluminosilicate, or aluminum solids (i.e., gibbsite and boehmite) could be the cause of excessive fouling of the crossflow or secondary filter at ARP. The authors conducted the tests by preparing slurries containing 6.6 M sodium Salt Batch 6 supernate, 2.5 g MST/L slurry, and varying concentrations of sodium oxalate, sodium aluminosilicate, and aluminum solids, processing the slurry through a bench-scale filter unit that contains a crossflow primary filter and a dead-end secondary filter, and measuring filter flux and transmembrane pressure as a function of time. Among the conclusions drwn from this work are the following: (1) All of the tests showed some evidence of fouling the secondary filter. This fouling could be from fine particles passing through the crossflow filter. (2) The sodium oxalate-containing feeds behaved differently from the sodium aluminosilicate- and gibbsite/boehmite-containing feeds.

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