Caustic-Side Solvent Extraction Research Articles

An Advanced Solvent for the Caustic-Side Solvent Extraction of Cesium from Nuclear Waste: Comparing Lipophilic Guanidines for Improved Hydrolytic Stability

ABSTRACT An advanced solvent has been developed for application in the extraction of cesium from alkaline nuclear waste. Four lipophilic guanidines have been compared to arrive at a more hydrolysis-resistant suppressor component of the solvent. Plans call for deployment of the Next-Generation Caustic Side-Solvent Extraction (NG-CSSX) process at industrial scale for separation of radioactive Cs-137 from the legacy salt waste stored at the US Department of Energy Savannah River Site (SRS). In the solvent used in NG-CSSX, an alkyl guanidine “suppressor” component facilitates efficient stripping of the loaded cesium from the solvent. However, as typical of guanidine compounds, the suppressor previously used in the NG-CSSX solvent, N,N‘,N’’-tri(3,7-dimethyloctyl)guanidine (TiDG), suffers from hydrolytic degradation under process conditions. Recently, the sterically hindered alkyl guanidine N,N’-dicyclohexyl-N’’-(10-nonadecyl)guanidine (DCNDG) has been found to offer 8 to 44 times greater hydrolysis resistance than TiDG. Here, the process performance characteristics of the NG-CSSX solvent incorporating DCNDG are determined in comparison with TiDG and two other guanidine suppressors. The solvent has been adjusted in density to accommodate use in the SRS Salt Waste Processing Facility and tested under aggressive bench-scale conditions intended to increase plant throughput. Experiments include the effect of ageing at normal and off-normal temperatures of the distribution of Cs+ through the NG-CSSX process, the fate and effects of guanidine degradation products, suppressor capacity, coalescence times for aqueous-solvent dispersions, tendency for emulsification, third-phase formation, and degree of protonation of the guanidine in stripping. While the chosen structures of the four compared guanidines mainly effect differences in stability, subtle differences in other system properties such as emulsion formation and suppressor capacity provide insight into the function of this important solvent component. In comparison with TiDG and two other guanidines, DCNDG provides greatly increased stability while not compromising the excellent functional properties of the NG-CSSX solvent.

Lipophilic Guanidine with Enhanced Stability for Use in Cesium Separation from Legacy High-Level Nuclear Waste

Guanidine functional groups are found in a variety of biologically active molecules and specialized molecules used in industrial applications but often suffer from instability because of their tendency to hydrolyze. Here, guanidines with differently structured alkyl groups were synthesized and their stabilities tested to determine if there is a relationship between the locus of branching in the alkyl groups attached to a guanidine and its hydrolytic stability under the hypothesis that increased steric hindrance to hydrolytic attack will increase stability. The guanidines examined in this work are pertinent to the next-generation caustic-side solvent extraction (NG-CSSX) process, which was developed for the removal of radioactive cesium from highly radioactive and complex alkaline solutions. In the NG-CSSX process, a lipophilic alkylguanidine solvent component is highly important to maintain the effectiveness of the Cs+ stripping of the loaded organic solvent putatively by sequestering extractable anions. Currently used for this purpose, N,N′,N″-tri(3,7-dimethyloctyl)guanidine (TiDG) succumbs to slow hydrolysis under process conditions. Guided by quantum chemical calculations, two new guanidines with large, sterically bulky alkyl substituents have been designed, synthesized, and tested. More sterically hindering alkyl groups were found to increase the stability of guanidines, particularly for the guanidine with branching closest to the guanidine N atom: N,N′-dicyclohexyl-N″-(10-nonadecyl)guanidine (DCNDG). DCNDG was found to have a stability toward hydrolysis 8 to 44 times greater than that of TiDG under different simulated NG-CSSX process conditions, making it one of the most hydrolysis-resistant guanidine molecules reported to date.

Thermal stability study of a new guanidine suppressor for the next-generation caustic-side solvent extraction process

ABSTRACTCesium stripping performance of thermally stressed solvent worsens slowly over time in batch tests of the Next-Generation Caustic-Side Solvent Extraction (NG-CSSX) process. NG-CSSX is currently used in full-scale equipment at the Savannah River Site for the selective removal of caesium from high-level salt waste. Recently, a new guanidine, N,N’,N”-tris(3,7-dimethyloctyl)guanidine (TiDG), was chosen for use as the suppressor, a lipophilic organic base needed for stripping, and the present study was undertaken to address the question of its stability. The NG-CSSX process solvent was evaluated for a period of three months under a variety of temperature and storage conditions. The performance of the solvent was tested at 30-day increments using an extraction, scrub, strip and extraction (ES2S3E) sequence. The results provide insights into the effects of storage and process conditions, the stripping behaviour of TiDG and the stability of the new solvent composition.

Determination of the Impact of Glycolate on Cs, Sr, and Actinide Separations in High-Level Waste

The Defense Waste Processing Facility (DWPF), responsible for vitrifying high-level waste (HLW) at the Savannah River Site (SRS), is planning to introduce glycolic acid as a reductant during chemical conditioning of the waste prior to vitrification. A portion of this glycolic acid may be recycled back to the tank farm as sodium glycolate. Since glycolate can serve as a complexing agent for a variety of metallic ions, recent studies were performed to examine the effect of glycolate on the removal of Sr and actinides by monosodium titanate (MST) and modified MST (mMST) and the separation of Cs by the caustic-side solvent extraction (CSSX) process.

Thermal Degradation of the Solvent Employed in the Next-Generation Caustic-Side Solvent Extraction Process and its Effect on the Extraction, Scrubbing, and Stripping of Cesium

As part of the ongoing development of the Next-Generation Caustic-Side Solvent Extraction (NG-CSSX) process, the thermal stability of its process solvent, the Next-Generation Caustic-Side Solvent (NGS) was investigated and shown to be adequate for industrial application. The solvent was thermally treated at 35°C over a period of 13 months whilst in dynamic contact with each of the aqueous phases of the current NG-CSSX process, namely SRS−15 (a highly caustic waste simulant), sodium hydroxide scrub solution (0.025 M), and boric acid strip solution (0.01 M). The effect of thermal treatment was evaluated by assessing batch extract/scrub/strip performance as a function of time, by monitoring the sodium extraction capacity of the solvent, and by analysis of the solvent using electrospray mass spectrometry. Current studies indicate that the NGS should be thermally robust without replenishment for a period of 7 months in the Modular Caustic-Side Solvent Extraction Unit (MCU), which has been treating waste using NG-CSSX since early 2014 at the Savannah River Site. The guanidine suppressor appears to be the solvent component most significantly impacted by thermal treatment of the solvent, showing significant degradation over extended operation.

Radiolytic Treatment of the Next-Generation Caustic-Side Solvent Extraction (NGS) Solvent and its Effect on the NGS Process

It is shown in this work that the solvent used in the Next Generation Caustic-Side Solvent Extraction (NGS) process can withstand a radiation dose well in excess of the dose it would receive in multiple years of treating legacy salt waste at the US Department of Energy Savannah River Site. The solvent was subjected to a maximum of 50 kGy of gamma radiation while in dynamic contact with each of the aqueous phases of the current NGS process, namely SRS−15 (a highly caustic waste simulant), sodium hydroxide scrub solution (0.025 M), and boric acid strip solution (0.01 M). Bench-top testing of irradiated solvent confirmed that irradiation has inconsequential impact on the extraction, scrubbing, and stripping performance of the solvent up to 13 times the estimated 0.73 kGy/y annual absorbed dose. Stripping performance is the most sensitive step to radiation, deteriorating more due to buildup of p-sec-butylphenol (SBP) and possibly other proton-ionizable products than to degradation of the guanidine suppressor, as shown by chemical analyses.

Performance Testing of the Next-Generation CSSX Solvent with Actual SRS Tank Waste

Efforts are underway to qualify the Next-Generation Solvent for the Caustic Side Solvent Extraction (CSSX) process. Researchers at multiple national laboratories have been involved in this effort. As part of the effort to qualify the solvent extraction system at the Savannah River Site (SRS), Savannah River National Laboratory (SRNL) researchers performed a number of tests at various scales. A series of batch equilibrium, or Extraction-Scrub-Strip (ESS), tests were conducted first. These tests used ∼30 mL of Next-Generation Solvent and either actual SRS tank waste, or waste simulant solutions. The results from these cesium mass transfer tests were used to predict solvent behavior under a number of conditions. For larger-scale testing, twelve stages of 2-cm (diameter) centrifugal contactors were assembled. This rack of contactors is structurally similar to one tested in 2001 during the demonstration of the baseline CSSX process. No issues were encountered during assembly and mechanical testing. A non-radiological test was performed using 35 L of cesium-spiked caustic waste simulant followed by a test with 39 L of actual tank waste. Test results are discussed, particularly those related to the effectiveness of extraction.

N, N′-Dicyclohexyl-N″-Isotridecylguanidine as Suppressor for the Next Generation Caustic Side Solvent Extraction (NG-CSSX) Process

The purity, concentration, and source of the N,N′-dicyclohexyl-N″-isotridecylguanidine (DCiTG) suppressor (guanidine) used in the NG-CSSX process were found to influence solvent performance. As the starting isotridecanol used in the preparation of DCiTG is comprised of a mixture of branched-chain aliphatic alcohols, varying in composition with manufacturer, the resulting DCiTG itself is a mixture. Thus, it is necessary to address how the solvent performance will be affected by the different preparations of the DCiTG solvent component. In this study, four preparations of DCiTG from three sources were analyzed and evaluated for purity and performance, both in the absence and presence of a deliberately added anionic surfactant impurity.

Characterization of Modified Monosodium Titanate – An Improved Sorbent for Strontium and Actinide Separations

High-level nuclear waste produced from fuel reprocessing operations at the Savannah River Site requires pretreatment to remove 134,137Cs, 90Sr, and alpha-emitting radionuclides (i.e., actinides) prior to disposal onsite as low level waste. An inorganic sorbent, monosodium titanate, is currently used to remove 90Sr and alpha-emitting radionuclides, while a caustic-side solvent extraction process is used for removing 134,137Cs. A new peroxotitanate material has recently been developed and has shown increased removal kinetics and capacity for 90Sr and alpha-emitting radionuclides compared to the current baseline material. This article describes recent results focused on further characterization of this material.

Development of an Improved Titanate-Based Sorbent for Strontium and Actinide Separations under Strongly Alkaline Conditions

High-level nuclear waste produced from fuel reprocessing operations at the Savannah River Site (SRS) requires pretreatment to remove 134,137Cs, 90Sr, and alpha-emitting radionuclides (i.e., actinides) prior to disposal onsite as low level waste. The separation processes at SRS include the sorption of 90Sr and alpha-emitting radionuclides onto monosodium titanate (MST) and caustic side solvent extraction of 137Cs. The MST and separated 137Cs is encapsulated along with the sludge fraction of high-level waste (HLW) into a borosilicate glass waste form for eventual entombment at a federal repository. The predominant alpha-emitting radionuclides in the highly alkaline waste solutions include plutonium isotopes 238Pu, 239Pu, and 240Pu; 237Np; and uranium isotopes, 235U and 238U. This article describes recent results evaluating the performance of an improved sodium titanate material that exhibits increased removal kinetics and capacity for 90Sr and alpha-emitting radionuclides compared to the current baseline material, MST.

Processing Macrobatch 2 at the Savannah River Site Integrated Salt Disposition Process (ISDP)

The Savannah River Site (SRS) is currently removing liquid radioactive waste from the tanks in its Tank Farm. To treat waste streams that are high in 137Cs, 90Sr, and/or actinides, SRS developed the Actinide Removal Process (ARP) and the Modular Caustic Side Solvent Extraction (CSSX) Unit. Collectively, these two processes make up the Integrated Salt Disposition Process (ISDP). The ARP part is responsible for the removal of strontium and actinides, while the MCU part is responsible for removing cesium. This paper discusses the qualification testing of the second batch of caustic waste that is being processed through ISDP currently. This paper also describes the tests conducted and compares results with current facility performance. The ARP contacts the salt solution with monosodium titanate (MST) to sorb strontium and select actinides. After MST contact, the resulting slurry is filtered to remove the MST (and sorbed strontium and actinides) and entrained sludge. The filtrate is transferred to the MCU for further treatment to remove cesium. The solid particulates removed by the filter are concentrated to ∼5 wt%, washed to reduce the sodium concentration, and transferred to the Defense Waste Processing Facility (DWPF) for vitrification. The CSSX process extracts the cesium from the radioactive waste using a customized solvent to produce a Decontaminated Salt Solution (DSS), then strips and concentrates the cesium from the solvent with dilute nitric acid. The DSS is incorporated in grout while the strip acid solution is transferred to DWPF for vitrification. In order to predict waste behavior, the Savannah River National Laboratory (SRNL) personnel performed tests using actual radioactive samples of the second waste batch – Macrobatch 2 – for processing prior to the start of the operation. Testing included MST sorption to remove strontium and actinides followed by CSSX batch contact tests to verify expected cesium mass removal and concentration. This paper describes the tests conducted and compares results from MCU facility operations. The results include strontium, plutonium, and cesium removal, cesium concentration, and organic entrainment and recovery data. Our work indicates that the bench scale tests are a conservative predictor of actual waste performance.

Robustness of the CSSX Process to Feed Variation: Efficient Cesium Removal from the High Potassium Wastes at Hanford

This contribution finds the Caustic‐Side Solvent Extraction (CSSX) process to be effective for the removal of cesium from the Hanford tank‐waste supernatant solutions. The Hanford waste types are more challenging than those at the Savannah River Site (SRS) in that they contain significantly higher levels of potassium, the chief competing ion in the extraction of cesium. By use of a computerized CSSX thermodynamic model, it was calculated that the higher levels of potassium depress the cesium distribution ratio (D Cs), as validated to within ±11% by the measurement of D Cs values on various Hanford waste‐simulant compositions. A simple analog model equation that can be readily applied in a spreadsheet for estimating the D Cs values for the varying waste compositions was developed and shown to yield nearly identical estimates as the computerized CSSX model. It is concluded from the batch distribution experiments, the physical‐property measurements, the equilibrium modeling, the flowsheet calculations, and the contactor sizing that the CSSX process as currently formulated for cesium removal from alkaline salt waste at the SRS is capable of treating similar Hanford tank feeds, albeit with more stages. For the most challenging Hanford waste composition tested, 31 stages would be required to provide a cesium decontamination factor (DF) of 5000 and a concentration factor (CF) of 2. Commercial contacting equipment with rotor diameters of 10 in. for extraction and 5 in. for stripping should have the capacity to meet throughput requirements, but testing will be required to confirm that the needed efficiency and hydraulic performance are actually obtainable. Markedly improved flowsheet performance was calculated based on experimental distribution ratios determined for an improved solvent formulation employing the more soluble cesium extractant BEHBCalixC6 used with alternative scrub and strip solutions, respectively 0.1 M NaOH and 0.010 M boric acid. The improved solvent and flowsheet can meet minimum requirements (DF = 5000 and CF = 2) with 15 stages or more ambitious goals (DF = 40,000 and CF = 15) with 19 stages. Thus, a modular CSSX application for the Hanford waste seems readily obtainable with further short‐term development.

Alternatives to Nitric Acid Stripping in the Caustic‐Side Solvent Extraction (CSSX) Process for Cesium Removal from Alkaline High‐Level Waste

Effective alternatives to nitric acid stripping in the Caustic‐Side Solvent Extraction (CSSX) solvent have been demonstrated in this work. The CSSX solvent employs calix[4]arene‐bis(tert‐octylbenzo‐18‐crown‐6) (BOBCalixC6) as the cesium extractant in a modified alkane diluent for decontamination of alkaline high‐level wastes, such as those found at the Savannah River and Hanford sites. Results reported in this paper support the idea that replacement of the nitrate anion by a much more hydrophilic anion like borate can substantially lower cesium distribution ratios on stripping. Without any other change in the CSSX flowsheet, however, the use of a boric acid stripping solution in place of the 1 mM nitric acid solution used in the CSSX process marginally, though perhaps still usefully, improves stripping. The less‐than‐expected improvement was explained by the carryover of nitrate from scrubbing into stripping. Accordingly, more effective stripping is obtained after a scrub of the solvent with 0.1 M sodium hydroxide. Functional alternatives to boric acid include sodium bicarbonate or cesium hydroxide as strip solutions. Profound stripping improvement is achieved when trioctylamine, one of the components of the CSSX solvent, is replaced with a commercial guanidine reagent (LIX® 79). The more basic guanidine affords greater latitude in the selection of aqueous conditions in that it becomes protonated even at mildly alkaline pH values. Under process‐relevant conditions, cesium distribution ratios on stripping are approximately 100‐fold lower compared with the current CSSX performance. The tolerance to surfactants is good, and the extraction properties of the solvent remain unchanged over three successive extract‐scrub‐strip cycles. From the point of view of compatibility with downstream processing, boric acid represents an attractive stripping agent, as it is also a potentially ideal feed for borosilicate vitrification of the separated 137Cs product stream. The possibilities for use of these results toward a dramatically better next‐generation CSSX process, possibly one employing the more soluble cesium extractant calix[4]arene‐bis(2‐ethylhexylbenzo‐18‐crown‐6) (BEHBCalixC6) are discussed.

Thermal and Spectroscopic Analyses of Caustic Side Solvent Extraction Solvent Contacted with 16 Molar and 8 Molar Nitric Acid

Thermal and spectroscopic analyses were performed on multiple layers formed when Caustic Side Solvent Extraction (CSSX) solvent was contacted with 16 M or 8 M nitric acid. A rapid (i.e., minutes to hours) chemical reaction occurs between the solvent and 16 M or 8 M nitric acid as evidenced by color changes and the detection of nitro groups in the infrared spectrum of the aged samples. Thermal analysis revealed that decomposition of the resulting mixture does not meet the definition of explosive or deflagrating material.

Full-Scale Testing of a Caustic Side Solvent Extraction System to Remove Cesium from Savannah River Site Radioactive Waste

Savannah River Site (SRS) personnel have completed construction and assembly of the Modular Caustic Side Solvent Extraction Unit (MCU) facility. Following assembly, they conducted testing to evaluate the ability of the process to remove non-radioactive cesium and to separate the aqueous and organic phases. They conducted tests at salt solution flow rates of 3.5, 6.0, and 8.5 gpm. During testing, the MCU Facility collected samples and submitted them to Savannah River National Laboratory (SRNL) personnel for analysis of cesium, Isopar® L, and modifier [1-(2,2,3,3-tetrafluoropropoxy)-3-(4-sec-butylphenoxy)-2-propanol]. SRNL personnel analyzed the aqueous samples for cesium by Inductively-Coupled Plasma Mass Spectroscopy (ICP-MS) and the solvent samples for cesium using a Parr Bomb digestion followed by ICP-MS. They analyzed aqueous samples for Isopar® L and modifier by gas chromatography (GC). The conclusions from the cesium analyses follow. The cesium in the feed samples measured 15.8 mg/L, in agreement with expectations. The decontamination factor measured 181–1580 at a salt solution flow rate of 3.5 gpm, 211–252 at a salt solution flow rate of 6.0 gpm, and 275–878 at a salt solution flow rate of 8.5 gpm. The concentration factor measured 11.0–11.1 at 3.5 gpm salt solution flow rate, 12.8–13.2 at 6.0 gpm salt solution flow rate, and 12.0–13.2 at 8.5 gpm salt solution flow rate. The organic carryover from the final extraction contactor (#7) varied between 22 and 710 mg/L Isopar® L. The organic carryover was less at the lowest flow rate. The organic carryover from the final strip contactor (#7) varied between 80 and 180 mg/L Isopar® L. The organic carryover in the Decontaminated Salt Solution Hold Tank (DSSHT) and the Strip Effluent Hold Tank (SEHT) was less than 10 mg/L Isopar® L, indicating good recovery of the solvent by the coalescers and decanters.

Scale-up of Caustic-Side Solvent Extraction Process for Removal of Cesium at Savannah River Site

The caustic-side solvent extraction process to remove radioactive cesium from Department of Energy salt waste was developed using small-scale annular centrifugal contactors. A scale-up factor of 750X is required. Testing with commercially available centrifugal contactors identified key design and operational parameters. Air flow is significant and the rotor bottom must remain covered to prevent surging flow. The Clean-in-Place feature was modified to allow hands-on maintenance. Flow rates and rotor speeds were optimized to reduce other-phase carryover. Startup, operation, and recovery methods were confirmed. Testing validated a Decontamination Factor > 12 and a strip Concentration Factor of 12 to 15.

Analysis of cesium extracting solvent using GCMS and HPLC

A high-level waste (HLW) remediation process scheduled to begin in 2007 at the Savannah River Site is the Modular Caustic Side Solvent Extraction (CSSX) Unit (MCU). The MCU will use a hydrocarbon solvent (diluent) containing a cesium extractant, a calix[4]arene compound, to extract radioactive cesium from caustic HLW. The resulting decontaminated HLW waste or raffinate will be processed into grout at the Saltstone Production Facility (SPF). The cesium containing CSSX stream will undergo washing with dilute nitric acid followed by stripping of the cesium nitrate into a very dilute nitric acid or the strip effluent stream and the CSSX solvent will be recycled. The Defense Waste Processing Facility (DWPF) will receive the strip effluent stream and immobilize the cesium into borosilicate glass. Excess CSSX solvent carryover from the MCU creates a potential flammability problem during DWPF processing. Bench-scale DWPF process testing was performed with simulated waste to determine the fate of the CSSX solvent components. A simple high performance liquid chromatography (HPLC) method was developed to identify the modifier (which is used to increase Cs extraction and extractant solubility) and extractant within the DWPF process. The diluent and triocytlamine (which is used to suppress impurity effect and ion-pair disassociation) were determined using gas chromatography mass spectroscopy (GCMS). To close the organic balance, two types of sample preparation methods were needed. One involved extracting aqueous samples with methylene chloride or hexane, and the second was capturing the off gas of the DWPF process using carbon tubes and rinsing the tubes with carbon disulfide for analysis. This paper addresses the development of the analytical methods and the bench-scale simulated waste study results.

Combined Extraction of Cesium and Strontium from Alkaline Nitrate Solutions

The combined extraction of cesium and strontium from caustic wastes can be achieved by adding a crown ether and a carboxylic acid to the Caustic‐Side Solvent Extraction (CSSX) solvent. The ligand 4,4′(5′)‐di(tert‐butyl)cyclohexano‐18‐crown‐6 and one of four different carboxylic acids were combined with the components of the CSSX solvent optimized for the extraction of cesium, allowing for the simultaneous extraction of cesium and strontium from alkaline nitrate media simulating alkaline high level wastes present at the U.S. Department of Energy Savannah River Site. Extraction and stripping experiments were conducted independently and exhibited adequate results for mimicking waste simulant processing through batch contacts. The promising results of these batch tests showed that the system could reasonably be tested on actual waste.

Strontium and Actinide Separations from High Level Nuclear Waste Solutions Using Monosodium Titanate 1. Simulant Testing

High‐level nuclear waste produced from fuel reprocessing operations at the Savannah River Site (SRS) requires pretreatment to remove 137Cs, 90Sr, and alpha‐emitting radionuclides (i.e., actinides) prior to disposal. Separation processes planned at SRS include caustic side solvent extraction, for 137Cs removal, and ion exchange/sorption of 90Sr and alpha‐emitting radionuclides with an inorganic material, monosodium titanate (MST). The predominant alpha‐emitting radionuclides in the highly alkaline waste solutions include plutonium isotopes 238Pu, 239Pu, and 240Pu. This paper provides a summary of data acquired to measure the performance of MST to remove strontium and actinides from simulated waste solutions. These tests evaluated the influence of ionic strength, temperature, solution composition, and the oxidation state of plutonium.

Cesium Removal from Savannah River Site Radioactive Waste using the Caustic-Side Solvent Extraction (CSSX) Process

:Researchers at the Savannah River National Laboratory successfully demonstrated the Caustic-Side Solvent Extraction (CSSX) process flow sheet using a 33-stage, 2-cm centrifugal contactor apparatus in two 24 h tests using actual high-level waste. The CSSX process for removal of cesium from alkaline solutions is the reference process for decontamination of high-level waste (HLW) at the Savannah River Site (SRS). The solvent consists of a calix[4]arene-crown-6 extractant, an alkyl-phenoxy alcohol modifier, and trioctylamine dissolved in an inert hydrocarbon matrix. Previously, we demonstrated the solvent extraction process with actual SRS HLW supernatant solution using a nonoptimized solvent formulation. Following that test, the solvent system was optimized to enhance extractant solubility in the diluent by increasing the modifier concentration. We now report results of two tests with the new and optimized solvent. The first test used a composite of supernatant solutions from two waste tanks, and the second test used a solution derived from dissolved salt cake. Test results showed that the CSSX process using the optimized solvent reduces 137Cs in HLW supernatant solutions to concentrations below the waste acceptance criterion of 45nCi/g for disposal as low-level waste (called “saltstone”). Waste decontamination factors as high as three million were achieved during testing. Test durations exceeded 24 h of uninterrupted operation and demonstrated hydraulic stability of the contactor array while operating with the optimized solvent. Carryover of organic solvent in aqueous streams (and aqueous in organic streams) was found to be less Ilian 1 %. The concentration factor (i.e., the ratio of the cesium concentration in the strip raftinate to the concentration in the waste) averaged approximately 13 during both tests, slightly below the process requirement of 15. Uncertainties in process flow-rate measurement and control prevented the test from achieving the target of 15.