Extraction of plutonium from acidic media using various phosphine oxides

Extraction, loading and stripping studies of Pu(IV) have been carried out using three phosphine oxides namely Cyanex^Ò-923 (cyn-923), Cyanex^Ò-925 (cyn-925) and TOPO in dodecane from nitric acid medium. All the three phosphine oxides have shown very high extraction of Pu. The order of extraction for Pu by these compounds is cyn-923 > TOPO - cyn-925. Loading of Pu (30.0 mg/l) in 3.0M HNO₃ was carried out using 5% solution of each of the phosphine oxides in dodecane. It was found that even at an organic to aqueous phase ratio of 1:10, the loading of Pu is >96%. From the loaded organic phase, Pu could be almost quantitatively stripped using 0.1 or 0.5M oxalic acid. The extraction of Pu(IV) with cyn-925 has also been carried out from HCl, HNO₃ or HClO₄ (0.5 to 9.1M). The species extracted into the cyn-925/dodecane phase from 3.0M HNO₃ or HCl media was found to be Pu(L)₄ ^.2 cyn-925 where L = NO₃ ^– or Cl^–. Similar species were observed to be formed when dodecane was replaced by xylene, chlorobenzene or o-dichlorobenzene.     

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% -activity originally present in the HLW, whereas with synthetic SBHLW the -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 SO ₄ ^2– (in the SBHLW) the complex ions AmSO ₄ ⁺ , UO₂SO₄, PuSO ₄ ²⁺ 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.     

Radiation chemistry of actinides

The paper presents data on the development of a mathematical model describing the radiation-chemical behaviour of plutonium in 3–6 mol/l HNO₃ with uranium present and absent. Using experimental data on the kinetics of Pu(VI) reduction and Pu(IV) oxidation, and comparing them with the calculated values, we succeeded in finding the previously unknown rate constants of many reactions between plutonium ions and HNO₃ radiolysis products, between plutonium ions and between plutonium ions and uranium ions. The mechanism of the radiation-chemical processes has been established.     

Purification and recovery of plutonium from uranium oxide obtained in the fast reactor fuel reprocessing plant using hydroxyurea as reductant

Spent fuel discharged from Fast Breeder Test Reactor (FBTR) in Kalpakkam is being reprocessed by modified plutonium uranium reduction extraction (PUREX) process using 30% TBP (tributylphosphate) as extractant in the presence of heavy normal paraffin (HNP) as diluent. Partitioning of uranium (U) and plutonium (Pu) is carried out using oxalate precipitation method. Uranium oxide product obtained by this method contains appreciable amount of plutonium which has to be recovered. Recovery of plutonium from this uranium oxide product is carried out by reducing Pu to inextractable Pu(III) using hydroxyurea (HU) and then uranium is extracted into 30% TBP. A small amount of Pu which is extracted in the organic phase is stripped back to aqueous phase by scrubbing with scrubbing agent containing 0.1 M HU in 4 M nitric acid. Similarly U and Pu are co-extracted into 30% TBP and then Pu is removed by scrubbing with 0.1 M HU in 4 M nitric acid. Further decontamination from Pu is obtained in the stripping stages. By this method Pu contamination in the uranium oxide is brought from 7300 ppm to 0.4–3 ppm (wt/wt). This uranium product obtained can be handled on table top.     

210Pb and210Po in tobacco-with a special focus on estimating the doses of210Po to man

Studies have been carried out on the solubility of Pu(III) oxalate by precipitation of Pu(III) oxalate from varying concentrations of HNO₃/HCl (0.5–2.0M) solutions and also by equilibrating freshly prepared Pu(III) oxalate with solutions containing varying concentrations of HNO₃/HCl, oxalic acid and ascorbic acid. Pu(III) solutions in HNO₃ and HCl media were prepared by reduction of Pu(IV) with ascorbic acid. 0.01–0.10M ascorbic acid concentration in the aqueous solution was maintained as holding reductant. The solubility of Pu(III) oxalate was found to be a minimum in 0.5M–1M HNO₃/HCl solutions containing 0.05M ascorbic acid and 0.2M excess oxalic acid in the supernatant.     

Extraction of uranium and plutonium from oxalate bearing solutions using phosphonic acid: Solvent extraction, extraction chromatography and infrared studies

Solvent extraction and extraction chromatography studies of uranium and plutonium from oxalate supernatant solutions were carried out using 2-ethyl hexyl-2-ethyl hexyl phosphonic acid (PC88A). Based on the distribution data, it was inferred that both the uranium and plutonium could be recovered satisfactorily from such a solution. These studies were found to be useful in optimising the appropriate concentration of PC88A, HNO₃, oxalic acid and temperature to recover more than 90% of plutonium from the large volumes of oxalate bearing waste solutions. Spectral characteristics of the extractant and its complexing behavior with U(VI) was also studied using IR & FTIR.     

Studies on the recovery of Pu(IV) from hydrochloric acid-oxalic acid solutions using an anion exchange method

Sorption of Pu(IV) from hydrochloric acid-oxalic acid solutions has been investigated using different anion exchangers, viz., Dowex-1X4, Amberlite XE-270 (MP) and Amberlyst A-26 (MP) for the recovery of plutonium from plutonium oxalate solutions. Distribution ratios of Pu(IV) for its sorption on these anion exchangers have been determined. The sorption of Pu(IV) from hydrochloric acid solutions decreases drastically in the presence of oxalic acid. However, addition of aluminium chloride enhances the sorption of plutonium in the presence of oxalic acid, indicating the feasibility of recovery of plutonium. Pu(IV) breakthrough capacities have been determined with a 10 ml resin bed of each of these anion exchangers at a flow rate of 60 ml per hour using a solution of Pu(IV) with the composition: 6M HCl+0.05M HNO₃+0.1M H₂C₂O₄+0.5M AlCl₃+100 mg.l^–1 Pu(IV). The 10% Pu(IV) breakthrough capacities for Dowex-1X4, Amberlite XE-270 (MP) and Amberlyst A-26 (MP) are 15.0, 8.9 and 6.2 g of Pu(IV) l^–1 of resin respectively.     

Separation of neptunium, plutonium, americium and curium from uranium with di-(2-ethylhexyl)-phosphoric acid (HDEHP) for radiometric and ICP-MS analysis

The possibility of using di-(2-ethylhexyl)-phosphoric acid (HDEHP) in solvent extraction for the separation of neptunium, plutonium, americium and curium from large amounts of uranium was studied. Neptunium, plutonium, americium and curium (as well as uranium) were extracted from HNO₃, whereafter americium and curium were back-extracted with 5M HNO₃. Thereafter was neptunium back-extracted in 1M HNO₃ containing hydroxylamine hydronitrate. Finally, plutonium was back-extracted in 3M HCl containing Ti(III). The method separates²³⁸Pu from²⁴¹Am for α-spectroscopy. For ICP-MS analysis, the interferences from²³⁸U are eliminated: tailing from²³⁸U, for analysis of²³⁷Np, and the interference of²³⁸UH⁺ for analysis of²³⁹Pu. The method has been used for the analysis of actinides in samples from a spent nuclear fuel leaching and radionuclide transport experiment.     

Tail-end purification of americium from plutonium loading effluents using a mixture of octyl (phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide and tri-N-butyl phosphate

The tail-end purification of Am from Pu loading effluents in 7.5M HNO₃ containing 160 mg l^–1 Am and 1.2 mg l^–1 Pu has been carried out. With 0.2M CMPO+1.2M TBP in dodecane as the extractant and stripping by 0.04M HNO₃+0.05M NaNO₂, the Pu level is brought down to 31.2 g l^–1. When the acidity was reduced to 4.2M HNO₃, one contact with 20% TLA/dodecane and subsequent extraction by a mixture of CMPO and TBP and stripping with 0.04M HNO₃+0.05M NaNO₂ gave Am samples without any detectable amounts of Pu. The recovery of Am was 90% by the first procedure and 98% by the second one.     

Separation of uranium from different uranium oxide matrices employing supercritical carbon dioxide extraction

Uranium from different uranium oxide matrices was extracted with tri-n-butyl phosphate–nitric acid (TBP–HNO₃) adduct using supercritical carbon dioxide (SC CO₂). While 30 min dissolution time at 323 K was sufficient for U₃O₈ and UO₂ powder, UO₂ granule (at 333 K) and crushed green pellet (at 353 K) required 40 min. Crushed sintered pellet required 60 min at 353 K for complete dissolution. Influence of various experimental parameters such as temperature, pressure, volume of TBP–HNO₃ adduct, acidity of nitric acid used for preparing TBP–HNO₃ adduct and extraction time on uranium extraction efficiency was also investigated. For UO₂ powder, temperature of 323 K, pressure of 15.2 MPa, 1 mL TBP–HNO₃ adduct, 10 M nitric acid and 30 min extraction time was found to be optimum. ~70% uranium extraction efficiency was obtained on extraction with SC CO₂ alone which increased to 90% with the addition of 2.5% TBP in SC CO₂ stream. Extraction efficiency was found to vary linearly with TBP percentage and nearly complete uranium extraction (~99%) was observed with 20% TBP. Nearly complete extraction was also achieved with addition of 2.5% thenoyltrifluoroacetylacetone (TTA) in methanol. The optimized procedure was extended to remove uranium from simulated tissue paper waste matrix smeared with uranium oxide solids.     

Recovery of plutonium from oxalate supernatant using 1-phenyl-3-methyl-4-benzoyl-5-pyrazolone

Summary Extraction of Pu(IV) from oxalate supernatant was carried out employing 1-phenyl-3-methyl-4-benzoyl-5-pyrazolone (PMBP) in xylene as extractant. The conditions for quantitative extraction were determined by the variation of ligand, oxalic acid and nitric acid concentration. Quantitative stripping was achieved using a mixture of 0.4M oxalic acid and 0.4M ammonium oxalate. Extraction of Pu(IV) from synthetic oxalate supernatant solution containing 3M nitric acid and 0.2M oxalic acid was investigated under various loading conditions employing 1-phenyl-3-methyl-4-benzoyl-5-pyrazolone in xylene as extractant. Under uranium loading conditions the Pu extraction decreased significantly while with increased Pu loading whereas the D_Pu value was influenced marginally. The effect of a redox reagent on Pu extraction was also investigated.     

Extraction of Am(III) from different acid media by di-2-ethyl hexyl phosphoric acid containing phosphorous pentoxide

The extraction of Am(III) from nitric, hydrochloric, oxalic, phosphoric and hydrofluoric acids was studied using 0.4F di-2-ethyl hexyl phosphoric acid (HDEHP) containing 0.1M phosphorous pentoxide (P₂O₅) in dodecane/xylene. The extraction with pure 0.4F HDEHP was found to be negligible from all the media studied. However, the presence of a small amount of P₂O₅ in it increased the extraction substantially. The distribution ratios of Am(III) obtained for HDEHP - P₂O₅ mixture 3M nitric acid containing different concentrations of oxalic acid/phosphoric acid/hydrofluoric acid are in the order of 200-250. The same for 3M hydrochloric acid is very high (800). These distribution ratios are sufficiently high for the quantitative extraction of Am(III) from all the acid media studied. Different reagents such as ammonium oxalate, sodium oxalate, oxalic acid, hydrofluoric acid, sodium carbonate and potassium sulphate were explored for the back extraction of Am(III) from 0.4F HDEHP + 0.1M P₂O₅ in dodecane/xylene. Of these, 0.35M ammonium oxalate and 1M sodium carbonate were found to be most suitable. The back extraction of Am(III) was also attempted with water and 1M H₂SO₄, HNO₃, HClO₄ and HCl solutions after allowing the extracted organics to degrade on its own. It was found that more than 90% of Am could be back extracted with these acids. Using this method more than 90% of Am(III) was recovered from nitric acid solutions containing calcium and fluoride ions.     

Titrimetric determination of uranium in the presence of plutonium in sulphuric acid medium

Present work summairzes a method for the estimation of uranium in the presence of plutonium involving the reduction of uranium to U/IV/ and plutonium to Pu/III/ by Zn/Hg/ followed by the selective oxidation of Pu/III/to Pu/IV/with HNO₃ catalyzed by molybdate in the presence of large sulphate concenration [5M H₂SO₄+1.5M /NH₄/₂SO₄]. The oxidation of U/IV/ by K₂Cr₂O₇ is then carried out in the presence of excess of Fe/III/ and Al/NO₃/₃ to a sharp potentiometric end point. R.S.D. obtained for 20 determinations of uranium /3–6 mg/ was 0.3% in the presence of 0.35 mg of plutonium. Larger quantity for plutonium was found to interfere.     

Use of organophosphorus extractants impregnated on silica gel for the extraction chromatographic separation of minor actinides from high level waste solutions

Silica-gel has been used as an inert support for the extraction chromatographic separation of actinides and lanthanides from HNO₃ and synthetic high level waste (HLW) solutions. Silica-gel was impregnated with tri-butyl phosphate (TBP), to yield STBP; 2-ethylhexyl phosphonic acid, mono 2-ethylhexyl ester (KSM-17, equivalent to PC-88A), SKSM; octyl(phenyl)-N,N-diisobutyl carbamoylmethylphosphine oxide (CMPO), SCMPO; and trialkylphosphine oxide (Cyanex-923), SCYN and sorption of Pu(IV), Am(III) and Eu(III) from HNO₃ solutions was studied batchwise. Several parameters, like time of equilibration, HNO₃ and Pu(IV) concentrations were varied. The uptake of Pu(IV) from 3.0M HNO₃ followed the order SCMPO>SCYN>SKSM>STBP. With increasing HNO₃ concentration, D _Pu increased up to 3.0M of HNO₃ for STBP, SKSM and SCMPO and then decreased. In the case of Am and Eu with SCMPO, the D values initially increased between 0.5 to 1.0M of HNO₃, remained constant up to 5.0M and then slightly decreased at 7.5M. Also, the effects of NaNO₃, Nd(III) and U(VI) concentrations on the uptake of Am(III) from HNO₃ solutions were evaluated. With increasing NaNO₃ concentration up to 3.0M, D _Am remained almost constant while it was observed that it decreases drastically by adding Nd(III) or U(VI). The uptake of Pu and Am from synthetic pressurized heavy water reactor high level waste (PHWR-HLW) in presence of high concentrations of uranium and after depleting the uranium content, and finally extraction chromatographic column separation of Pu and Am from U-depleted synthetic PHWR-HLW have been carried out. Using SCMPO, high sorption of Pu, Am and U was obtained from the U-depleted HLW solution. These metal ions were subsequently eluted using various reagents. The sorption results of the metal ions on silica-gel impregnated with several phosphorus based extractants have been compared. The uptake of Am, Pu and rare earths by SCMPO has been compared with those where CMPO was sorbed on Chromosorb-102, Amberchrom CG-71 and styrene divinylbenzene copolymer immobilized in porous silica particles.     

Alpha radiolytic degradation of <Emphasis Type="Italic">N,N</Emphasis>-dihexyl octanamide

The primary purpose of this study was to understand the alpha radiolytic degradation behavior of N,N-dihexyl octanamide (DHOA) vis a vis tributyl phosphate (TBP) solutions in n-dodecane under plutonium loading conditions. These studies were carried out as a function of dose on different Pu loaded samples (containing 0.002-10 g/L Pu) from 4 M HNO₃ medium. These Pu loaded solutions were evaluated for stripping behavior by contacting with 0.5 M NH₂OH at 0.5 M HNO₃ solutions. Organic phase analysis was carried out by gas chromatography (GC) and by visible spectrophotometry. These studies clearly indicated that Pu stripping becomes difficult with increased dose in the case of TBP system. On the other hand, no such problem was observed in DHOA system during stripping of plutonium, thereby indicating that DHOA is a promising candidate for the reprocessing of high burn up Pu rich spent fuels.     

A routine method for the determination of plutonium in urine with a low detection limit

A radiochemical method for the determination of plutonium in urine is described. The steps involved are a) co-precipitation of plutonium, b) wet ashing, c) hydrolysis, d) extraction from 2M HNO₃ into capillary polypropylene columns coated with tri-n-octyl phosphineoxide 0.5M in toluene, and e) back-extraction of plutonium from the organic phase, f) electroplating onto stainless steel disks and spectrometry, since plutonium is extracted together with small amounts of uranium naturally occurring in urine. High quality deposits for spectrometry are obtained because iron interference is eliminated before back-extraction. The radiochemical recovery of²³⁹Pu is 55.6±7.5% and the detection limit is 1.0 mBq per liter of urine.     

Studies on purification of plutonium from silver ions

Precipitation and solvent extraction methods have been investigated for the purification of plutonium from silver from the solution generated during oxidative dissolution of plutonium oxide using Ag(II) ions. Initial experiments have been carried out using thorium as representative of plutonium. Selecting the optimum conditions, the experiments were repeated with plutonium. The results revealed that Pu can be purified from silver ions either by precipitating silver as silver chloride or silver metal followed by Pu(IV) oxalate precipitation or by selective extraction of Pu(IV) into 20% Aliquat-336 or 30% TBP.     

Extraction of actinides and fission products by octyl(phenyl)-N,N-diisobutylcarbamoylmethyl-phosphine oxide from nitric acid media

Extraction of promethium(III), uranium(VI), plutonium(IV), americium(III), zirconium(IV), ruthenium(III), iron(III) and palladium(II) has been carried out with a mixture of octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) and tributyl phosphate (TBP) in dodecane. The effects of nitric acid, TBP and CMPO concentrations on the extraction of these metal ions have been studied. The nature of the species of the above metal ions extracted into the organic phase has been suggested.     

Preparation, characterization and application of thorium oxalate for sorption of plutonium and americium from aqueous solution

Synthetic inorganic exchangers exhibit good thermal and radiation stability. Thorium oxalate precipitate shows potential for co-precipitation of plutonium and americium from oxalate supernatant generated during plutonium oxalate precipitation. In the present study, efforts were made to prepare thorium oxalate precipitate to be used for column operation. Distribution ratios were determined to optimize conditions for sorption of plutonium and americium on thorium oxalate from nitric acid + oxalic acid solutions with composition similar to that of oxalate supernatant. Column experiments were also performed to evaluate the sorption capacity of thorium oxalate for plutonium and americium from the same medium. The result showed that, thorium oxalate prepared in 1.75M HNO₃ at 70 °C is suitable for column operations. These studies showed that plutonium and americium could be simultaneously removed from aqueous solutions with composition similar to plutonium oxalate waste using glass column packed with thorium oxalate and these nuclides could be recovered by eluting with 3M HNO₃.     

Ion exchange studies of U(VI), Th(IV), Pu(IV) and Eu(III) on a macroreticular resin in TBP-Shell Sol-T solutions

Ion exchange studies of uranium(VI), thorium(IV), plutonium(IV) and europium(III) ions on a macroreticular cation exchange resin, Amberlyst A-15, from solutions of 30% and 5% TBP—Shell Sol-T have been carried out. The metal ions were extracted into TBP Shell Sol-T phase from 8M NH₄NO₃ at different nitric acid concentrations. Ion exchange distribution ratios as a function of organic phase acidity of 30% and 5% TBP have been computed. Separation factors computed from the observed Kd values are plotted as a function of organic phase acidity.