V. Gopalakrishnan

Partitioning of actinides from high-level waste streams of purex process using mixtures of CMPO and TBP in dodecane

Abstract The extraction of actinides from high active aqueous raffinate waste (HAW) as well as high-level waste (HLW) solutions arising from Purex processing of thermal reactor fuels has been studied using a mixture of octyl(phenyl)-N N-diisobutylcarbamoyl-methylphosphine oxide (CMPO) and TBP in dodecane. The results on the extraction and stripping of actinides, lanthanides, and other fission products are discussed. Optimum conditions are proposed for the efficient recovery of residual actinides from HAW and HLW streams by CMPO extraction followed by their selective stripping with suitable reagents. Experiments on the extraction and separation of actinides and lanthanides by CMPO in the presence of TBP in dodecane have also been carried out with U(VI) and Nd(III) to arrive at the limiting conditions for avoiding third-phase formation.

Separation of carrier-free 90Y from high level waste by supported liquid membrane using KSM-17

A generator system has been developed for the preparation of carrier-free 90Y from 90Sr present in the high level waste (HLW) of the Purex process by employing a supported liquid membrane (SLM) using 2-ethylhexyl-2-ethylhexyl phosphonic acid (KSM-17 equivalent to PC 88A) supported on a polytetrafluoro ethylene (PTFE) membrane. When uranium depleted Purex HLW at appropriate acidity is passed sequentially through octyl (phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) sorbed on chromosorb-102 (abbreviated as CAC) and Zeolite AR1 (synthetic mordenite) columns, all the trivalent, tetravalent and hexavalent metal ions and monovalent 137Cs ions are sorbed. After adjusting to pH 2 with NaOH the resulting effluent is used as feed in a single stage membrane cell partitioned with a PTFE membrane impregnated with KSM-17 and having a feed and receiver compartment with 5.0 ml capacity each. The receiver compartment was filled with a 0.5M HNO3 or 0.5M HCl stripping solution. 90Y alone is preferentially transported across the membrane leaving behind all the impurities viz. 90Sr, 125Sb, 106Ru, 106Rh, etc. in the feed compartment. This technique can yield 90Y in mCi levels in a pure and carrier-free form for medical applications. The feed can be reused repeatedly after allowing for 90Y buildup.

Recovery of actinides extracted by Truex solvent from high level waste using complexing agents

This work deals with the batch studies on stripping of actinides extracted by a mixture octyl(phenyl)-N,N-diisobutyl-carbamoylmethylphosphine oxide (CMPO) and tri-n-butyl phosphate (TBP) in n-dodecane (Truex solvent) from simulated high level waste (HLW) solution. The stripping of americium and plutonium from acid-bearing CMPO-TBP mixture is carried out using a mixture of weak acid, weak base and complexing agent as strippant. A mixture of formic acid, hydrazine hydrate and citric acid appeared to be best suited for efficient stripping of americium and plutonium. With appropriate modifications in the concentration of individual constitutents, this strippant can be used for the recovery of actinides from loaded Truex solvent with any acid content.

Separation of Carrier-Free 90Y from High Level Waste by Extraction Chromatographic Technique Using 2-Ethylhexyl-2-ethylhexyl Phosphonic Acid (KSM-17)

An extraction chromatographic technique has been developed for the separation of carrier-free 90Y from the 90Sr present in the high level waste (HLW) of the Purex process. When a Purex HLW solution in 2–3 M HNO3 is passed through a CMPO-Chromosorb-102 (CAC) column, all the trivalent, tetravalent, and hexavalent ions are sorbed. The effluent from this experiment, after adjusting the pH to 2 with NaOH, was passed through a 2-ethylhexyl-2-ethylhexyl phosphonic acid (KSM-17)-Chromosorb-102 (KSMC) extraction chromatographic column where only 90Y was sorbed. All the other ions (90Sr, 137Cs, 125Sb, 106Ru, 106Rh, etc.) were washed off with dilute HNO3 (pH 2), and carrier-free 90Y was eluted with 0.5 M HNO3. This technique can yield 90Y in mCi levels in pure form for medical applications. The 90Sr can be used repeatedly after allowing for 90Y buildup.

Extraction and extraction chromatographic separation of minor actinides from sulphate bearing high level waste solutions using CMPO

Bench-Scale studies on the partitioning and recovery of minoractinides from the actual and synthetic sulphate-bearing high level waste (SBHLW) solutions have been carried out by giving two contacts with 30% TBP to deplete uranium content followed by four contacts with 0.2M CMPO+1.2M TBP in dodecane. The acidity of the SBHLW solutions was about 0.3M. In the case of actual SBHLW, the final raffinate contained about 0.4% α-activity originally present in the HLW, whereas with synthetic SBHLW the α-activity was reduced to the background level.144Ce is extracted almost quantitative in the CMPO phase,106Ru about 12% and137Cs is practically not extracted at all. The extraction chromatographic column studies with synthetic SBHLW (aftertwo TBP contacts) has shown that large volume of waste solutions could be passed through the column without break-through of actinide metal ions. Using 0.04M HNO3>99% Am(III) and rare earths could be eluted/stripped. Similarly >99% Pu(IV) and U(VI) could be eluted.stripped using 0.01M oxalic acid and 0.25M sodium carbonate, respectively. In the presence of 0.16M SO42− (in the SBHLW) the complex ions AmSO4+, UO2SO4, PuSO42+ and Pu(SO4)2 were formed in the aqueous phase but the species extracted into the organic phase (CMPO+TBP) were only the nitrato complexes Am(NO3)3·3CMPO, UO2(NO3)2·2CMPO and Pu(NO3)4·2CMPO. A scheme for the recovery of minor actinides from SBHLW solution with two contacts of 30% TBP followed by either solvent extraction or extraction chromatographic techniques has been proposed.

STUDIES ON THE PARTITIONING OF ACTINIDES FROM HIGH LEVEL WASTE USING A MIXTURE OF HDEHP AND CMPO AS EXTRACTANT

The paper describes the extraction and stripping behavior of actinides and lanthanides viz. americium, plutonium, uranium, cerium, and europium using a mixture of di-(2-ethylhexyl)phosphoric acid (HDEHP) and octyl(phenyl)-N,N-diisobutylcarbamoyl- methylphosphine oxide (CMPO) in n-paraffin as extractant. The extractant mixture combines the advantages of both the extractants for the partitioning of actinides from PUREX High Level Waste (HLW) solution. The extraction studies have been carried out from nitric acid medium as well as with simulated uranium-lean PUREX-HLW solution as relevant to Pressurized Heavy Water Reactor (PHWR) fuel reprocessing. The extracted actinides are stripped using a mixture of diethylenetriaminepentaacetic acid (DTPA), formic acid and hydrazine hydrate, leaving lanthanides in the organic phase. The lanthanides are subsequently stripped using a mixture of DTPA and sodium carbonate. The behavior of cesium, strontium, ruthenium, and zirconium is also studied using this mixed extractant.

EXTRACTION CHROMATOGRAPHIC SEPARATION OF PROMETHIUM FROM HIGH ACTIVE WASTE SOLUTIONS OF PUREX ORIGIN

ABSTRACT An extraction chromatographic procedure for the separation of 147Pm from High Active Waste solutions of Purex process has been developed. Octy1(pheny1)-N,N-diisobutylcarbamoylmethy1phosphine oxide(CMPO) and 2-ethylhexy1-2-ethylhexylphosphonic acid (KSM-17), both sorbed separately on an inert support(chromosorb-102) have been sequentially employed for this purpose. In the CMPO column, the rare earths and the trivalent actinides are sorbed together with uranium, plutonium and traces of few other fission products. The elution of this column with 0.04 M HNO3 gives an eluate containing trivalent actinides and lanthanides. This solution, after adjusting the pH to 2.0, is used as feed for the second extraction chromatographic column based on KSM-17. All the trivalent metal ions are sorbed on the column leaving the trace impurities in the effluent. Fractional elution of the metal ions from this column is carried out with nitric acid of varying concentrations, At 0.09 M HNO3, the pure beta emitting fracti...

Recovery of Neptunium from Highly Radioactive Waste Solutions of Purex Origin Using Tributyl Phosphate

Abstract The present work deals with the extraction of neptunium into the TBP/dodecane phase under conditions relevant to highly radioactive waste solutions, along with uranium and plutonium, by oxidizing it to the hexavalent state using 0.01 M K2Cr2O7 and subsequently recovering it by selective stripping. Three types of simulated HLW solutions, namely sulfate-bearing (SB, in ∼0.3 M HNO3) and non sulfate wastes originating from the reprocessing of fuels from pressurised heavy water reactors (PHWR) and fast breeder reactors (FBR) (both in 3.0 M HNO3), have been employed in this study. Very high extraction of U(VI), Np(VI), and Pu(VI) was obtained from PHWR and FBR-HLW solutions, whereas extraction was less but reasonably high from the SB-HLW solution. The uptake of cerium at tracer level concentrations in the millimolar range (encountered in HLW solutions) and from the simulated HLW solutions containing 0.01 M K2Cr2O7 by 30% TBP has shown that its extraction takes place only at tracer level concentrations ...

Recovery of actinides extracted by Truex solvent from high level waste using complexing agents: II. Counter-current studies

These studies are an extension of carlier work on the recovery of actinides extracted by Truex solvent from simulated high level waste solution, with a mixture of weak acid, weak base and complexing agent used as a strippant. The effectiveness of the proposed strippant, consisting of formic acid, hydrazine hydrate and citric acid, is tested in a counter-current mode using mixer-settler in the present studies. The studies show that near quantitative recovery of americium and plutonium is feasible from acid-bearing Truex solvent with no reflux (reextration) of activity. Use of this strippant reduces considerably the generation of secondary waste.

Separation and Recovery of Uranium, Neptunium, and Plutonium from High Level Waste Using Tributyl Phosphate: Countercurrent Studies with Simulated Waste Solution

ABSTRACT The present work deals with countercurrent extraction studies on the partitioning of uranium, neptunium, and plutonium using 30% tributyl phosphate (TBP) from simulated high level waste solution generated during reprocessing of spent uranium fuel from pressurized heavy water reactors. The oxidation states of neptunium and plutonium were adjusted either by 0.01 M potassium dichromate or 0.01 M dioxovanadium ion. Neptunium and plutonium, extracted in the TBP phase, were stripped together using a mixture containing 0.05 M ascorbic acid and 0.25 M hydrogen peroxide in 2.0 M nitric acid solution. Although dioxovanadium ion is more effective for proper adjustment of the oxidation states of plutonium and neptunium, subsequent recovery of these actinides from loaded TBP is better if potassium dichromate is used for the valency adjustment. Results of the stagewise analysis of extraction and stripping of actinides using mixer-settlers are presented.