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

Abstract

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.     

干法后处理废盐中活泼裂片元素的净化工艺研究进展

随着快堆研究的快速发展,干法后处理工艺流程也逐渐成为了研究重点,熔盐电解干法后处理是未来先进核燃料循环系统的核心环节和关键技术。氯化锂-氯化钾共晶盐是干法后处理工艺中最常用的熔盐体系。为了提取乏燃料中的锕系、镧系和铯、锶元素,需要对熔盐进行长时间的电解。在锕系分离提取过程中,镧系和铯、锶等活泼裂片元素在熔盐中不断积累,不仅会改变熔盐体系的理化性质,还将影响后续锕系产品的净化效果。为实现溶剂盐复用,使放射性废物最小化,需定期对废熔盐中的镧系和铯、锶等活泼裂片元素进行净化处理。对干法后处理氯化锂-氯化钾废熔盐中镧系和铯、锶等活泼裂片元素采取的净化工艺,包括熔盐萃取法、熔盐电解法、沉淀法、区域精炼法等工艺的原理、特点和研发进展进行了综述和比较分析,讨论了上述工艺中为实现溶剂盐复用、减少放射性废物产生对废熔盐中的镧系和铯、锶等活泼裂片元素的净化效果。指出了中国废盐净化将围绕实现稀土资源利用最大化、保护环境、最大程度上减少废物的排放开展相关方向的研究。     

Strontium and Actinide Separations from High Level Nuclear Waste Solutions Using Monosodium Titanate 2. Actual Waste Testing

Abstract

Pretreatment processes at the Savannah River Site will separate ⁹⁰Sr, alpha‐emitting and radionuclides (i.e., actinides) and ¹³⁷Cs prior to disposal of the high‐level nuclear waste. Separation of ⁹⁰Sr and alpha‐emitting radionuclides occurs by ion exchange/adsorption using an inorganic material, monosodium titanate (MST). Previously reported testing with simulants indicates that the MST exhibits high selectivity for strontium and actinides in high ionic strength and strongly alkaline salt solutions. This paper provides a summary of data acquired to measure the performance of MST to remove strontium and actinides from actual waste solutions. These tests evaluated the effects of ionic strength, mixing, elevated alpha activities, and multiple contacts of the waste with MST. Tests also provided confirmation that MST performs well at much larger laboratory scales (300 – 700 times larger) and exhibits little affinity for desorption of strontium and plutonium during washing.     

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.     

γ‐Radiation Effects on the Performance of HCCD‐PEG for Cs and Sr Extraction

Abstract

Cobalt dicarbollide and polyethylene glycol in phenyltrifluoromethyl sulfone (HCCD/PEG in FS‐13) is currently under consideration for use in the process‐scale selective extraction of fission product cesium and strontium from acidic radioactive solutions. While the Cs and Sr solvent extraction efficiency of this formulation has been previously characterized, this solvent will be exposed to high radiation doses during use, and has not been adequately investigated for radiation stability. Here, HCCD/PEG was γ‐irradiated to various absorbed doses, to a maximum of 432 kGy, using ⁶⁰Co. Irradiations were performed for the neat organic phase, and also for the organic phase in contact with 1 M‐nitric acid mixed by air sparging. Post‐irradiation solvent extraction measurements showed that Cs distribution ratios were unaffected; however, strontium distribution ratios decreased with the absorbed dose under both conditions. The decrease in the extraction efficiency for strontium was greater when in contact with the aqueous phase. The stripping performance was not affected. A mechanism, based on reaction with the products of direct diluent radiolysis, is proposed to explain the decreases in the strontium extraction efficiency.     

Development and Testing of Srex Flowsheets for Treatment of Idaho Chemical Processing Plant Sodium-Bearing Waste Using Centrifugal Contactors

Abstract

Laboratory experimentation has indicated that the SREX process is effective for partitioning ⁹⁰Sr from acidic radioactive waste solutions located at the Idaho Chemical Processing Plant. A baseline flowsheet has been proposed for the treatment of sodiumbearing waste (SBW) which includes extraction of strontium from liquid SBW into the SREX solvent (0.15 M 4′,4′ (5′)-di-(tert-butyldicyclohexo)-18-crown-6 and 1.2 M TBP in Isopar L®), a 0.01 M nitric acid strip section to back-extract components from the loaded solvent, and a 2.0 M HNO₃ solvent acidification section to equilibrate the solvent with HNO₃ prior to recycle to the extraction section. The flowsheet was designed to provide a decontamination factor (DF) of >10³ which will reduce the ⁹⁰Sr activity in the waste solution to below the NRC Class A LLW limit of 0.04 Ci ⁹⁰Sr/m³. SREX flowsheet testing was performed using sixteen stages of 5.5-cm diameter centrifugal contactors. The behavior of stable Sr and other components which are potentially extracted by the SREX solvent were evaluated. Specifically, the behavior of the matrix components including Pb, K, Hg, Na, Ca, Zr, and Fe was studied. The described flowsheet achieved 99.98% Sr removal (DF=4250) with one cycle of SREX. Potassium and Zr were partially extracted into the SREX solvent with 35% and 21%, respectively, exiting in the strip product. Sodium, Ca, and Fe were essentially inextractable. Lead was determined to extract and accumulate in the SREX solvent and in the strip section. As a result, a Pb precipitate formed in the strip stages of the contactors. Mercury was also determined to extract and accumulate in the SREX solvent.     

Actinide Partitioning with a Modified TODGA Solvent: Counter-Current Extraction Studies with Simulated High Level Waste

Actinide partitioning studies with improved N,N,N',N'-tetraoctyl diglycolamide (TODGA) solvent (0.05 M TODGA + 5% iso-decanol in n-dodecane) has been explored in order to achieve better decontamination from fission products. The distribution behavior of various metal ions, viz. Am, Pu, U, Eu, Sr, Pd, Cs, Tc, Fe, and Mo with improved TODGA solvent was investigated. Lower concentration of TODGA (0.05 M as compared to previously proposed 0.1 M or 0.2 M) exhibited required extraction properties for actinide partitioning from pressurized heavy water reactor high level waste (PHWR-HLW). Counter-current extraction studies with simulated PHWR-HLW spiked with different radio-tracers (viz. ²⁴¹Am, ¹⁵²Eu, ¹³⁷Cs, ^85,89Sr, ⁵⁹Fe, ¹⁰⁶Ru, ¹⁰⁹Pd, ⁹⁵Zr, and ⁹⁹Mo) suggested that > 99.9% of the trivalent actinides and lanthanides could be extracted in six stages and stripped in four stages. The decontamination factor for various fission products with respect to Am was: 3918 (Cs), 2990 (Sr), 1150 (Zr), 1407 (Ru), 1185 (Pd), and 3250 (Mo). The counter-current extraction studies with the irradiated solvent (500 kGy) reflected a significant amount of Mo extraction.     

Review Article: The Effects of Radiation Chemistry on Solvent Extraction 3: A Review of Actinide and Lanthanide Extraction

The partitioning of the long‐lived α‐emitters and the high‐yield fission products from dissolved used nuclear fuel is a key component of processes envisioned for the safe recycling of used nuclear fuel and the disposition of high‐level waste. These future processes will likely be based on aqueous solvent‐extraction technologies for light‐water reactor fuel and consist of four main components for the sequential separation of uranium, fission products, group trivalent actinides, and lanthanides, and then trivalent actinides from lanthanides. Since the solvent systems will be in contact with highly radioactive solutions, they must be robust toward radiolytic degradation in an irradiated mixed organic/aqueous acidic environment, with the formation of only benign degradation products. Therefore, an understanding of their radiation chemistry is important to the design of a practical system. In the first paper in this series, we reviewed the radiation chemistry of irradiated aqueous nitric acid and the tributyl phosphate ligand for uranium extraction in the first step of these extractions. In the second, we reviewed the radiation chemistry of the ligands proposed for use in the extraction of cesium and strontium fission products. Here, we review the radiation chemistry of the ligands that might be used for the group extraction of the lanthanides and actinides. This includes traditional organophosphorus reagents such as CMPO and HDEHP, as well as novel reagents such as the amides and diamides currently being investigated.     

Efficient co-extraction of strontium and cesium from nitric acid medium by mixtures of di-tert-butylcyclohexano-18-crown-6 and 1,3-di(2-propoxy)calix[4]arene-crown-6 in n-octanol

The co-extraction of strontium and cesium from nitric acid medium by di-tert-butylcyclohexano-18-crown-6 (DtBuCH18C6) and 1,3-di(2-propoxy)calix[4]arene-crown-6 (iPr-C[4]C-6) in n-octanol was studied. The effects of contact time, nitric acid concentration, extractant concentration and temperature on the co-extraction behavior were systematically investigated. Effective extraction of the two metals was achieved under a variety of conditions. The co-extraction from a simulated high-level liquid waste (HLLW) was also conducted, and strontium and cesium could be selectively extracted in the presence of a large number of other metals. Results in this work illustrate the feasibility of partitioning radioactive strontium and cesium simultaneously from HLLW by a mixture of DtBuCH18C6 and iPr-C[4]C-6 in n-octanol.     

An Advanced TALSPEAK Concept for Separating Minor Actinides. Part 2. Flowsheet Test with Actinide-spiked Simulant

An Advanced TALSPEAK (trivalent actinide–lanthanide separations by phosphorus-reagent extraction from aqueous complexes) counter-current flowsheet test was demonstrated using a simulated feed spiked with radionuclides in annular centrifugal contactors. A solvent comprising 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester (HEH[EHP] or PC88A) in n-dodecane was used to extract trivalent lanthanides away from the trivalent actinides Am³⁺ and Cm³⁺, which were preferentially complexed in a citrate-buffered aqueous phase with N-(2-hydroxyethyl)ethylenediamine-N,N´,N´-triacetic acid (HEDTA). In a 24-stage demonstration test, the trivalent actinides were efficiently separated from the trivalent lanthanides with decontamination factors >1000, demonstrating the excellent performance of the chemical system. Clean actinide and lanthanide product fractions and spent solvent with very low contaminations were obtained. The results of the process test are presented and discussed.     

Combined Extraction of Cesium and Strontium from Alkaline Nitrate Solutions

Abstract

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.     

The Use of 24-Crown-8's in the Solvent Extraction of CsNO3 and Sr(NO3)2

Abstract

The feasibility of a solvent extraction process for removing strontium and cesium from acidic high activity nuclear waste is shown. Both strontium and cesium can be extracted from an aqueous HNO₃ phase containing the metal nitrates into an organic phase containing kerosene or CCl₄ as a diluent and complexing agents dissolved in the diluent. The most promising results obtained thus far have required the use of a mixture of three metal complexing agents: tributyl phosphate, di-2-ethylhexyl phosphoric acid, and 4,4′(5′)-di-tert-butylbenzo-24-crown-8. The highest distribution coefficients obtained (organic/aqueous) were 1.45 ± 0.05 for Cs⁺ and 200 for Sr²⁺. The extraction is reversible and is strongly dependent on the pH of the aqueous phase. The metal can be removed from the organic phase by lowering the pH to 1, while raising the pH above 3 causes the metal to return to the organic phase. The utility of this extraction technique for nuclear processing will depend on the radiation stability of the complexing agents and the degree of selectivity obtained when extracting strontium and cesium from mixed fission products.     

Solvent Extraction of Strontium and Cesium: A Review of Recent Progress

Radioactive isotopes ¹³⁷Cs and ⁹⁰Sr, two significant fission products that are usually carried into High Level Waste (HLW) during spent nuclear fuel reprocessing are suggested to be removed from HLW in order to reduce the volume of HLW and then make nuclear energy more clean and sustainable. A variety of separation techniques, including solvent extraction, have been developed for the removal of ¹³⁷Cs and ⁹⁰Sr from HLW. Among those developed separation techniques, solvent extraction is more applicable and promising, particularly for acidic HLW. This article reviews the scientific progress as well as application developments of the solvent extraction method for the separation of strontium and cesium from HLW in the last decade.     

DIAMEX Counter‐Current Extraction Process for Recovery of Trivalent Actinides from Simulated High Active Concentrate

Abstract

The partitioning of trivalent actinides was demonstrated with a new version of the French DIAMEX (DIAMide EXtraction) process. A continuous counter‐current experiment using a 16‐stage centrifugal extractor battery was tested using 1 mol/L N,N′-dimethyl‐N,N′-dioctyl‐hexylethoxy‐malonamide (DMDOHEMA) in TPH as the extractant. A high active concentrate (HAC), obtained after concentration and denitration of a high active raffinate (HAR) with a concentration factor of 10, was used as a feed. Based on results from cold and hot batch extraction experiments and computer code calculations, a flowsheet was developed and a full test was carried out using a simulated HAC solution spiked with radionuclides (²⁴¹Am, ²⁴⁴Cm, ¹⁵²Eu, and ¹³⁴Cs). In the DIAMEX process, five extraction stages were sufficient to obtain Am and Cm (feed/raffinate) greater than 5000 and for the coextracted lanthanides decontamination factors between 1100 and 4500. Co‐extraction of zirconium, molybdenum, and palladium was prevented by using oxalic acid and HEDTA. The back extraction comprising 4 stages was also efficient and the recoveries of actinides were greater than 99.8%, which can be further improved by a minor process flowsheet optimisation. The experimental steady‐state concentration profiles of important solutes were determined and compared with model calculations and good agreement was generally obtained.     

Recovery of 137Cs from Laboratory Waste using Solvent Extraction with Sodium Tetraphenylboron (TPB)

The paper describes a method for the recovery of ¹³⁷Cs from an aqueous radioactive laboratory waste solution containing ¹³⁷Cs (2 µg/mL) in the presence of high concentration of Na⁺ using solvent extraction technique. The method comprises of adjustment of pH to the acidic range (pH = 2), contacting the aqueous radioactive solution with sodium tetraphenylboron (TPB) in nitrobenzene, whereby ¹³⁷Cs binds with tetraphenylboron anions and gets separated. Results of this investigation indicate that ¹³⁷Cs could be efficiently and selectively extracted from an aqueous solution media containing high concentration of Na⁺ under mildly acidic pH into an organic phase and back extracted with small volume of 3 M HNO₃, thus enabling concentration. The proposed method was successfully applied in real samples.     

Experimental Verification of Caustic‐Side Solvent Extraction for Removal of Cesium from Tank Waste

Abstract

A caustic‐side solvent extraction (CSSX) process was developed to remove Cs from Savannah River Site (SRS) high‐level waste. The CSSX process was verified in a series of flowsheet tests at Argonne National Laboratory (ANL) in a minicontactor (2‐cm centrifugal contactor) using simulant. The CSSX solvent, which was developed at Oak Ridge National Laboratory (ORNL), consists of a calixarene‐crown ether as the extractant, an alkyl aryl polyether as the modifier, trioctylamine as the suppressant, and Isopar®L as the diluent. For Cs removal from the SRS tank waste, the key process goals are that: (1) Cs is removed from the waste with a decontamination factor greater than 40,000 and (2) the recovered Cs is concentrated by a factor of 15 in dilute nitric acid. In the flowsheet verification tests, the objectives were to: (1) prove that these process goals could be met; (2) demonstrate that they could be maintained over a period of several days as the CSSX solvent is recycled; and (3) verify that the process goals could still be met after the solvent composition was adjusted. The change in composition eliminated the possibility that the calixarene‐crown ether could precipitate from the solvent. The process goals were met for each of the verification tests. The results of these tests, which are summarized here, show that the CSSX process is a very effective way to remove Cs from caustic‐side waste.     

Demonstration of a TODGA‐Based Continuous Counter‐Current Extraction Process for the Partitioning of Actinides from a Simulated PUREX Raffinate,Part II: Centrifugal Contactor Runs

Abstract

The efficiency of the partitioning of trivalent actinides from a PUREX raffinate is demonstrated with a TODGA+TBP extractant mixture dissolved in an industrial aliphatic solvent TPH. Based on the results of cold and hot batch extraction studies and with the aid of computer code calculations, a continuous counter‐current process is developed and two flowsheets are tested using miniature centrifugal contactors. The feed solution used is a synthetic PUREX raffinate, spiked with ²⁴¹Am, ²⁴⁴Cm, ²⁵²Cf, ¹⁵²Eu, and ¹³⁴Cs. More than 99.9% of the trivalent actinides and lanthanides are extracted and back‐extracted and very high decontamination factors are obtained for most fission products. The co‐extraction of zirconium, molybdenum, and palladium is prevented using oxalic acid and HEDTA. However, 10% of ruthenium is extracted and only 3% is back‐extracted using diluted nitric acid. The experimental steady‐state concentration profiles of important solutes are determined and compared with model calculations and good agreement is generally obtained.     

Flowsheet Modeling and Testing of Pseudohydroxide Extraction from Aqueous Sodium Hydroxide Solutions with 3,5‐di‐tert‐Butylphenol in Isopar® L Modified with Exxal® 8

Abstract

A conceptual counter‐current process flowsheet was developed for sodium hydroxide recovery from alkaline solutions via pseudohydroxide extraction (PHE). PHE relies on a simple sodium ion/proton exchange mechanism at elevated pH using a weak organic acid extractant. The contact of the sodium‐loaded organic phase with water results in the reconstitution of the extractant in the organic phase and sodium hydroxide in the aqueous phase. In this work, the 3,5‐di‐tert‐butylphenol (35‐DTBP) cation exchanger was used in the Isopar^® L diluent modified with isooctyl alcohol Exxal^® 8. Equilibrium isotherms determined for PHE from pure sodium hydroxide solutions and simulated radioactive waste leachate were used to develop a semi‐empirical model that could be used for designing PHE process flowsheets. Using this model, a conceptual PHE flowsheet was developed for recovering NaOH from solutions generated by caustic leaching of radioactive tank sludges. The flowsheet consists of the extraction, scrub, and strip processes, each employing four equilibrium stages. The modeling of this flowsheet indicates 97% recovery of the sodium hydroxide from the waste leachate feed solution. An experimental demonstration, performed with a simulated radioactive waste leachate using batch contacts in a co‐current analog of the counter‐current flowsheet, confirmed the potential for practical application of PHE technology.