Demonstration of Caustic-Side Solvent Extraction with Savannah River Site High-Level Waste PDF Download

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Languages : en
Pages : 5

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This report details the results of analyses for organic compounds and trace components in various process streams based on samples from the 48-hour radioactive waste test.

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Languages : en
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Researchers successfully demonstrated the chemistry and process equipment of the Caustic-Side Solvent Extraction (CSSX) flowsheet for the decontamination of high level waste using a 33-stage, 2-cm centrifugal contactor apparatus at the Savannah River Technology Center. This represents the first CSSX process demonstration using Savannah River Site (SRS) high level waste. Three tests lasting 6, 12, and 48 hours processed simulated average SRS waste, simulated Tank 37H/44F composite waste, and Tank 37H/44F high level waste, respectively.

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Languages : en
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This paper describes a demonstration of a solvent extraction process for removal of 137Cs from alkaline high level radioactive waste (HLW) solutions from the Savannah River Site (SRS) tank farm. The process employed a calix[4]arene-crown-6 extractant dissolved in an inert hydrocarbon matrix, along with an alkylphenoxy alcohol modifier, and an additional additive, trioctylamine, to improve stripping performance and mitigate the effects of organic anions present in the feed stream. Testing included washing and recycle of the solvent. Tests conducted with two non-radioactive simulants of SRS tank waste verified proper hydraulic operation prior to the radioactive experiments. An additional 12-hour demonstration used simulated SRS tank waste spiked with 137Cs to confirm proper hydraulic operation, decontamination and volume reduction factors following installation of the equipment in a shielded cell facility. Finally, a 48-hour demonstration was performed employing a composite waste from SRS tank farms. Approximately 106 liters of waste (1.5E+09 d/min/mL 137Cs activity) was processed through the test unit. Average 137Cs decontamination factors for the decontaminated waste streams in all tests exceeded the target value of 40,000. Concentration factors in the strip section agreed to within 5 percent of the target value of 15.

Author: M. A. Norato
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Languages : en
Pages : 5

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Researchers successfully demonstrated the chemistry of the Caustic-Side Solvent Extraction (CSSX) flow sheet with optimized solvent. This represents the third such process demonstration using actual Savannah River Site (SRS) high level waste (HLW). The present test differed from previous studies in the use of radioactive waste derived from Tank 37H dissolved salt cake, as opposed to supernate solutions used in previous demonstrations.

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Researchers successfully demonstrated the chemistry and process equipment of the Caustic-Side Solvent Extraction (CSSX) flowsheet using MaxCalix for the decontamination of high level waste (HLW). The demonstration was completed using a 12-stage, 2-cm centrifugal contactor apparatus at the Savannah River National Laboratory (SRNL). This represents the first CSSX process demonstration of the MaxCalix solvent system with Savannah River Site (SRS) HLW. Two tests lasting 24 and 27 hours processed non-radioactive simulated Tank 49H waste and actual Tank 49H HLW, respectively. A solvent extraction system for removal of cesium from alkaline solutions was developed utilizing a novel solvent invented at the Oak Ridge National Laboratory (ORNL). This solvent consists of a calix[4]arene-crown-6 extractant dissolved in an inert hydrocarbon matrix. A modifier is added to the solvent to enhance the extraction power of the calixarene and to prevent the formation of a third phase. An additional additive is used to improve stripping performance and to mitigate the effects of any surfactants present in the feed stream. The process that deploys this solvent system is known as Caustic Side Solvent Extraction (CSSX). The solvent system has been deployed at the Savannah River Site (SRS) in the Modular CSSX Unit (MCU) since 2008.

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Languages : en
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Researchers successfully demonstrated the chemistry and process equipment of the Caustic-Side Solvent Extraction (CSSX) flowsheet using MaxCalix for the decontamination of high level waste (HLW). The demonstration was completed using a 12-stage, 2-cm centrifugal contactor apparatus at the Savannah River National Laboratory (SRNL). This represents the first CSSX process demonstration of the MaxCalix solvent system with Savannah River Site (SRS) HLW. Two tests lasting 24 and 27 hours processed non-radioactive simulated Tank 49H waste and actual Tank 49H HLW, respectively. Conclusions from this work include the following. The CSSX process is capable of reducing 137Cs in high level radioactive waste by a factor of more than 40,000 using five extraction, two scrub, and five strip stages. Tests demonstrated extraction and strip section stage efficiencies of greater than 93% for the Tank 49H waste test and greater than 88% for the simulant waste test. During a test with HLW, researchers processed 39 liters of Tank 49H solution and the waste raffinate had an average decontamination factor (DF) of 6.78E+04, with a maximum of 1.08E+05. A simulant waste solution (≈34.5 liters) with an initial Cs concentration of 83.1 mg/L was processed and had an average DF greater than 5.9E+03, with a maximum DF of greater than 6.6E+03. The difference may be attributable to differences in contactor stage efficiencies. Test results showed the solvent can be stripped of cesium and recycled for ≈25 solvent turnovers without the occurrence of any measurable solvent degradation or negative effects from minor components. Based on the performance of the 12-stage 2-cm apparatus with the Tank 49H HLW, the projected DF for MCU with seven extraction, two scrub, and seven strip stages operating at a nominal efficiency of 90% is ≈388,000. At 95% stage efficiency, the DF in MCU would be ≈3.2 million. Carryover of organic solvent in aqueous streams (and aqueous in organic streams) was less than 0.1% when processing Tank 49H HLW. The entrained solvent concentration measured in the decontaminated salt solution (DSS) was as much as ≈140 mg/L, although that value may be overstated by as much as 50% due to modifier solubility in the DSS. The entrained solvent concentration was measured in the strip effluent (SE) and the results are pending. A steady-state concentration factor (CF) of 15.9 was achieved with Tank 49H HLW. Cesium distribution ratios [D(Cs)] were measured with non-radioactive Tank 49H waste simulant and actual Tank 49H waste. Below is a comparison of D(Cs) values of ESS and 2-cm tests. Batch Extraction-Strip-Scrub (ESS) tests yielded D(Cs) values for extraction of ≈81-88 for tests with Tank 49H waste and waste simulant. The results from the 2-cm contactor tests were in agreement with values of 58-92 for the Tank 49H HLW test and 54-83 for the simulant waste test. These values are consistent with the reference D(Cs) for extraction of ≈60. In tests with Tank 49H waste and waste simulant, batch ESS tests measured D(Cs) values for the two scrub stages as ≈3.5-5.0 for the first scrub stage and ≈1.0-3.0 for the second scrub stage. In the Tank 49H test, the D(Cs) values for the 2-cm test were far from the ESS values. A D(Cs) value of 161 was measured for the first scrub stage and 10.8 for the second scrub stage. The data suggest that the scrub stage is not operating as effectively as intended. For the simulant test, a D(Cs) value of 1.9 was measured for the first scrub stage; the sample from the second scrub stage was compromised. Measurements of the pH of all stage samples for the Tank 49H test showed that the pH for extraction and scrub stages was 14 and the pH for the strip stages was ≈7. It is expected that the pH of the second scrub stage would be ≈12-13. Batch ESS tests measured D(Cs) values for the strip stages to be ≈0.002-0.010. A high value in Strip No.3 of a test with simulant solution has been attributed to issues associated with the limits of detection for the analytical method. In the 2-cm contactor tests, the first four strip stages of the Tank 49H waste test and all five strip stages in the simulant waste test had higher values than the ESS tests. Only the fifth strip stage D(Cs) value of the Tank 49H waste test matched that of the ESS tests. It is speculated that the less-than-optimal performance of the strip section is caused by inefficiencies in the scrub section. Because strip is sensitive to pH, the elevated pH value in the second scrub stage may be the cause of strip performance. In spite of the D(Cs) values obtained in the scrub and strip sections, testing showed that the solvent system is robust. Average DFs for the process far exceeded targets even though the scrub and strip stages did not function optimally. Correction of the issue in the scrub and strip stages is expected to yield even higher waste DFs.

Author: Lætitia H. Delmau
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Category : Chemical engineering
Languages : en
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Author: JR. J. F. Birdwell
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Languages : en
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This report presents the results of a conceptual design of a solvent extraction process for the selective removal of {sup 137}Cs from high-level radioactive waste currently stored in underground tanks at the U.S. Department of Energy's Savannah River Site (SRS). This study establishes the need for and feasibility of deploying a simplified version of the Caustic-Side Solvent Extraction (CSSX) process; cost/benefit ratios ranging from 33 to 55 strongly support the considered deployment. Based on projected compositions, 18 million gallons of dissolved salt cake waste has been identified as having {sup 137}Cs concentrations that are substantially lower than the worst-case design basis for the CSSX system that is to be deployed as part of the Salt Waste Processing Facility (SWPF) but that does not meet the waste acceptance criteria for immobilization as grout in the Saltstone Manufacturing and Disposal Facility at SRS. Absent deployment of an alternative cesium removal process, this material will require treatment in the SWPF CSSX system, even though the cesium decontamination factor required is far less than that provided by that system. A conceptual design of a CSSX processing system designed for rapid deployment and having reduced cesium decontamination factor capability has been performed. The proposed accelerated-deployment CSSX system (CSSX-A) has been designed to have a processing rate of 3 million gallons per year, assuming 90% availability. At a more conservative availability of 75% (reflecting the novelty of the process), the annual processing capacity is 2.5 million gallons. The primary component of the process is a 20-stage cascade of centrifugal solvent extraction contactors. The decontamination and concentration factors are 40 and 15, respectively. The solvent, scrub, strip, and wash solutions are to have the same compositions as those planned for the SWPF CSSX system. As in the SWPF CSSX system, the solvent and scrub flow rates are equal. The system is designed to facilitate remote operation and direct maintenance. Two general deployment concepts were considered: (1) deployment in an existing but unused SRS facility and (2) deployment in transportable containers. Deployment in three transportable containers was selected as the preferred option, based on concerns regarding facility availability (due to competition from other processing alternatives) and decontamination and renovation costs. A risk assessment identified environmental, safety, and health issues that exist. These concerns have been addressed in the conceptual design by inclusion of mitigating system features. Due to the highly developed state of CSSX technology, only a few technical issues remain unresolved; however, none of these issues have the potential to make the technology unviable. Recommended development tasks that need to be performed to address technical uncertainties are discussed in this report. Deployment of the proposed CSSX-A system provides significant qualitative and quantitative benefits. The qualitative benefits include (1) verification of full-scale contactor performance under CSSX conditions that will support SWPF CSSX design and deployment; (2) development of design, fabrication, and installation experience bases that will be at least partially applicable to the SWPF CSSX system; and (3) availability of the CSSX-A system as a means of providing contactor-based solvent extraction system operating experience to SWPF CSSX operating personnel. Estimates of fixed capital investment, development costs, and annual operating cost for SRS deployment of the CSSX-A system (in mid-2003 dollars) are $9,165,199, $2,734,801, and $2,108,820, respectively. When the economics of the CSSX-A system are compared with those of the baseline SWPF CSSX system, benefit-to-cost ratios ranging from 20 to 47 are obtained. The benefits in the cost/benefit comparison arise from expedited tank closure and reduced engineering, construction, and operating costs for the SWPF CSSX system. No significant impediments to deployment were determined in the reported analysis, and where technical uncertainties were identified, development tasks to mitigate them are indicated. It is recommended that deployment of the CSSX-A system be pursued in a timely manner in order to derive the greatest possible cost and accelerated treatment benefits.

Author: National Research Council
Publisher: National Academies Press
ISBN: 0309071941
Category : Science
Languages : en
Pages : 154

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The Second World War introduced the world to nuclear weapons and their consequences. Behind the scene of these nuclear weapons and an aspect of their consequences is radioactive waste. Radioactive waste has varying degrees of harmfulness and poses a problem when it comes to storage and disposal. Radioactive waste is usually kept below ground in varying containers, which depend on how radioactive the waste it. High-level radioactive waste (HLW) can be stored in underground carbon-steel tanks. However, radioactive waste must also be further immobilized to ensure our safety. There are several sites in the United States where high-level radioactive waste (HLW) are stored; including the Savannah River Site (SRS), established in 1950 to produce plutonium and tritium isotopes for defense purposes. In order to further immobilize the radioactive waste at this site an in-tank precipitation (ITP) process is utilized. Through this method, the sludge portion of the tank wastes is being removed and immobilized in borosilicate glass for eventual disposal in a geological repository. As a result, a highly alkaline salt, present in both liquid and solid forms, is produced. The salt contains cesium, strontium, actinides such as plutonium and neptunium, and other radionuclides. But is this the best method? The National Research Council (NRC) has empanelled a committee, at the request of the U.S. Department of Energy (DOE), to provide an independent technical review of alternatives to the discontinued in-tank precipitation (ITP) process for treating the HLW stored in tanks at the SRS. Alternatives for High-Level Waste Salt Processing at the Savannah RIver Site summarizes the finding of the committee which sought to answer 4 questions including: "Was an appropriately comprehensive set of cesium partitioning alternatives identified and are there other alternatives that should be explored?" and "Are there significant barriers to the implementation of any of the preferred alternatives, taking into account their state of development and their ability to be integrated into the existing SRS HLW system?"

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Languages : en
Pages : 5

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Researchers at the Savannah River Technology Center (SRTC) successfully demonstrated the Caustic-Side Solvent Extraction (CSSX) process flow sheet using a 33-stage, 2-cm centrifugal contactor apparatus in two 24-hour tests using actual high level waste. Previously, we demonstrated the solvent extraction process with actual SRS HLW supernatant solution using a non-optimized 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.