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

The sorption and migration of radiostrontium in a calcareous soil from Yu Zhong county of Gansu province (China) was studied using batch and column experiments. Sorption isotherms, breakthrough curves and concentration profiles for the untreated soil and the soil treated to remove CaCO₃ were determined, respectively. It was found that radiostrontium is a relatively mobile nuclide in calcareous soil and removal of CaCO₃ from the soil slightly increases the retention ability for radiostrontium. The breakthrough curves were fitted to the analytical solution of a one-dimensional convection-dispersion transport model that assumes a reversible linear isotherm. Good agreement was obtained between the measured and predicted concentration profiles.     

Recovery of molybdenum by solvent extraction from simulated high level liquid waste

Journal of Radioanalytical and Nuclear Chemistry - In this study, simple and a rapid solvent extraction method is investigated to recover Mo(VI) from simulated HLLW. The (NH4)6Mo7O24·4H2O...     

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.     

Characterization of high level waste for minor actinides by chemical separation and alpha spectrometry

The high level waste (HLW) generated from the reprocessing of the spent fuel of pressurized heavy water reactor has been characterized for the minor actinides. The radiation dose of the waste solution was reduced by radiochemical separation of cesium from HLW by solvent extraction with chlorinated cobalt dicarbollide dissolved in 20% nitrobenzene in xylene. Minor actinides (Np, Pu, Am, Cm) in the high level waste were assayed by alpha spectrometry following radiochemical separation. The gross alpha activity determined by liquid scintillation agrees well (within 10%) with the cumulative quantities of actinides determined by alpha spectrometry.     

Scaling-up experience of actinides partitioning from intermediate level high salted alkaline waste streams of Purex process using Versatic-10 extractant

Partitioning of minor alpha-emitting actinides, especially U, Pu and Am from medium active alkaline waste is possible from intermediate level liquid wastes (ILLW) produced during spent fuel reprocessing following Purex process. This paper deals with the efficient removal of alpha-activity from ILLW by solvent extraction process. Counter current batch extraction with O/A ratio 2:1 as well as multistage mixer settler has demonstrated that most of the alpha-activity was removed from the alkaline effluents using 20% Versatic-10 (V-10) in dodecane after giving 3 to 4 contacts, thus converting alkaline waste as non-alpha waste. Under the optimum conditions (pH 9.0-9.5 and VA-10), both Pu(IV) and Am(III) are highly extractable whereas U(VI) is relatively poorly extracted. To assess the applicability of this process for regular treatment of the waste, a feasibility study on pilot plant scale using six stage mixer settler was operated to treat the ILLW. The results indicated that almost >99.90% alpha-emitting actinides are removed. Dilute nitric acid (0.5M HNO₃) served as the most efficient strippant for all these actinides. This facilitate an easy regeneration of the extractant which can be recycled. This method is useful in obtaining alpha-free wastes and had positive impact on ease and safety aspects during subsequent waste treatment and long term storage.     

Recovery of actinides from acidic waste solutions generated in research facilities using adsorption and extraction techniques

Journal of Radioanalytical and Nuclear Chemistry - In the present study various separation methods were investigated for the recovery of actinides from aqueous waste solutions. Extraction behavior...     

Separation of trivalent actinides from lanthanides by solvent extraction

A method is presented for separating the trivalent actinides, mainly Am and Cm, from trivalent lanthanides by the use of only two solvent extractants. The first solvent removes the heavy lanthanides, leaving the Am, Cm and the lighterlanthanides; the second removes the Am and Cm. Because additional complexing agents are not required, waste-disposal and corrosion problems are reduced. Overall separation factors may be as high as several thousand for the separation of Am and Cm from lanthanides in the fission waste products from reactor fuel processing.     

Sulphate complexing of some trivalent actinides

The sulphate complexing of the trivalent actinides Pu(III), Am(III) and Cm(III) in aqueous solutions was studied by the dinonylnaphthalene sulphonic acid extraction method, at ionic strengths of 1 and 2, at 25°C and the respective stability constant values obtained are reported. The extraction of the actinide ions from sulphuric acid into amines indicated the existence of their anionic sulphate complex species.     

Gross alpha-activity estimation in simulated high level waste after actinides partitioning by solid state nuclear track detection

The estimation of low level alpha activity is difficult in waste samples containing large concentration of salts and beta–gamma activity. In the present study, the feasibility of gross alpha-activity measurement for simulated high level waste (SHLW) in solution medium by alpha-track registration technique has been attempted. The results showed that it is possible to use this technique for gross alpha-activity estimation of ~200 Bq/mL in solution medium with a precision and accuracy of ~30%. The importance of measuring 200 Bq/mL alpha activity in SHLW solutions is that this value corresponds to about 4,000 Bq/g activity in the solid medium which is the safe disposable limit. The advantage of this method over other methods is that it is not sensitive to beta–gamma emitters and salts and is very simple and inexpensive.     

高放废液中分离回收镧系-锕系元素萃取剂的研究进展

概述了近十几年来国内外有关用于高放废液镧系-锕系元素萃取分离中新萃取剂的研究进展,并对酰胺类、硫代磷类、硫代吡唑啉酮类等新萃取剂的设计筛选等进行了讨论。     

Recovery of cesium from intermediate level liquid radioactive waste

The antioxidant Lowinox 22M46 (Naftonox 22M46) were used for the extraction of cesium from intermediate level liquid radioactive wastes.     

Separation of cesium from intermediate level liquid radioactive waste by solvent extraction with antioxidants

The intermediate level liquid radioactive wastes (RAW) isussed from nuclear power plants have high salt contents ca 200 g·dm^–3, the pH of liquid RAW being 12.5–13.7. A convenient method for separation of cesium under these conditions is solvent extraction with substituted phenols. For this purpose weere tested antioxidants produced in Czechoslovakia: AO 2246 [2,2-methylene-bis-(4-methyl-6-tertbutyl)phenol]; AO 4 [2-tertbutyl-4-(2-phenylpropyl)phenol]; AO 4K [2,6-ditertbuty-4-methylphenol]; AO 301 [2,2-methylene-bis-(4-{2-phenylpropyl}-6-tert-butyl)phenol]; and one antioxidant imporoted from Japan—NOCRAC 2246. This antioxidant is equivalent to AO 2246. After the first experiment it was found that the extraction efficiency for antioxidants AO 4 and Ao 301 is very low and the following experiments were made with AO 2246 (NOCRAC 2246) and AO 4K. Some effects on extracton as, pH of water phase, influence of diluent, influence of concentration of antioxidants, extraction time, were studied. The best results gave antioxidant NOCRAC 2246 in nitrobenzene, the extraction efficiency was 92.3% with pH 13.23.     

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.     

Studies on fluoride complexing of hexavalent actinides using a fluoride ion selective electrode

Complex formation between actinide(VI) and fluoride ions in aqueous solutions has been investigated using a fluoride ion selective electrode (F-ISE). As fairly high acidity was used to suppress hydrolysis of the actinide(VI) ions, significant liquid junction potentials (E_j) existed in the systems. An iterative procedure was developed for computing free hydrogen ion concentration [H⁺], as it could not be measured directly, using data obtained with F-ISE. E_j values were estimated from known [H⁺] and the stability constants of fluoride complexes of actinide(VI) ions were calculated following KING and GALLAGHER's method using a computer program. The stability constants were found to follow the order U(VI)>Np(VI)>Pu(VI).     

Biosorption of actinides from dilute waste actinide solution by egg-shell membrane

Removal of radioactive elements from the effluent and waste aqueous solutions is an important problem. In previous laboratory batch experiments, hen egg-shell membrane (ESM) was stable as an insoluble protein and was very capable of binding heavy metal ions from aqueous solution. Batch laboratory pH profile, time dependency, and capacity experiments were performed to determine the binding of uranium (U) and thorium (Th) to ESM. Batch pH profile experiments indicated that the optimum pH for binding these actinides was approx 6.0 (U) or 3.0 (Th). The adsorption isotherms were developed at pH 5.0 (U) or 3.0 (Th) at 25°C, and the adsorption equilibrium data fitted both Langmuir and Freundlich models. The maximum uptakes by the Langmuir model were about 240 mg U/g and 60 mg Th/g dry weight ESM. In addition, their adsorption capacities increased as salt concentration increased. ESM could also accumulate uranium from dilute aqueous solution by adjusting to the optimum pH. These results showed that ESM was effective for removing actinides from solution and would be useful in filtration technology to remove actinides from aqueous solution. S.-I. Ishikawa is a research fellow at the Japan Society for the Promotion of Science.     

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.     

Studies on the partitioning of actinides from high level liquid waste solution employing supported liquid membrane with Cyanex-923 as carrier

The present paper deals with the studies on the partitioning of actinides from high level liquid waste solution of PUREX origin employing supported liquid membrane technique. The process uses solution of Cyanex-923 in n-dodecane as a carrier with poly tetra fluoro ethylene support and a mixture of citric acid, formic acid and hydrazine hydrate as a receiving phase. Transport studies are carried out for ²⁴¹Am under different experimental conditions to optimize the transport parameters such as feed acidity, carrier concentration, effect of uranium, trivalent metal ion and salt concentration in the feed. Studies indicated good transport of actinides across the membrane from nitric acid medium. Under the optimized conditions the transport of ²⁴¹Am is studied from a uranium depleted synthetic PHWR-HLLW and finally the technique has been used for the partitioning of alpha emitters from an actual research reactor-HLLW. High concentration of uranium in the feed is found to retard the transport of americium, suggesting the need for prior removal of uranium from the waste.     

Removal of actinides and fission products activity from intermediate alkaline waste using inorganic exchangers

Hydrated iron oxide or amorphous-Fe₂O₃·3.5 H₂O (HFeO), hydrated titanium oxide (HTiO) and hydrated thorium oxide (HThO) were synthesized and their applicability for the decontamination of intermediate level liquid wastes (ILLW) was tested. The sorption of a few actinides like plutonium and americium on HFeO, ¹³⁷Cs and ¹⁰⁶Ru on HTiO and ⁹⁰Sr on HThO was investigated as a function of pH, time and loading capacity of the hydrous oxide with metal ions. The influence of the total dissolved salt content was also monitored. Some of these parameters influenced the sorption behavior significantly. The radiation stability of these inorganic sorbents were studied by irradiating them up to 48 Mrad. Adsorbed actinides and fission products were successfully eluted from HFeO and from the mix-bed of HTiO and HThO by 0.5M nitric acid.The authors wish to thank Shri R. D. Changarani, Chief Superintendent NRG Facilities and Shri P. K. Dey, Head FRD for their valuable advice and constant support.     

Removal of actinides and fission products activity from intermediate alkaline waste using inorganic exchangers

Hydrated iron oxide or amorphous-Fe₂O₃·3.5 H₂O (HFeO), hydrated titanium oxide (HTiO) and hydrated thorium oxide (HThO) were synthesized and their applicability for the decontamination of intermediate level liquid wastes (ILLW) was tested. The sorption of a few actinides like plutonium and americium on HFeO, ¹³⁷Cs and ¹⁰⁶Ru on HTiO and ⁹⁰Sr on HThO was investigated as a function of pH, time and loading capacity of the hydrous oxide with metal ions. The influence of the total dissolved salt content was also monitored. Some of these parameters influenced the sorption behavior significantly. The radiation stability of these inorganic sorbents were studied by irradiating them up to 48 Mrad. Adsorbed actinides and fission products were successfully eluted from HFeO and from the mix-bed of HTiO and HThO by 0.5M nitric acid.The authors wish to thank Shri R. D. Changarani, Chief Superintendent NRG Facilities and Shri P. K. Dey, Head FRD for their valuable advice and constant support.     

Recovery of uranium from uranium refining waste water by using immobilized persimmon tannin

Some attempts were made to examine the practical conditions for uranium recovery from uranium refining waste water. The adsorbent was highly effective in recovering uranium. The uranium adsorption was affected by pH, temperature, and uranium concentration of the uranium refining waste water. The adsorbent also recovered uranium effectively in column system. It aquires better mechanical properties and can be used repeatedly in the uranium adsorption-desorption cycles.