New approaches to reprocessing of oxide nuclear fuel

Dissolution of UO₂, U₃O₈, and solid solutions of actinides in UO₂ in subacid aqueous solutions (pH 0.9–1.4) of Fe(III) nitrate was studied. Complete dissolution of the oxides is attained at a molar ratio of ferric nitrate to uranium of 1.6. During this process actinides pass into the solution in the form of U(VI), Np(V), Pu(III), and Am(III). In the solutions obtained U(VI) is stable both at room temperature and at elevated temperatures (60 °C), and at high U concentrations (up to 300 mg mL^?1). Behavior of fission products corresponding to spent nuclear fuel of a WWER-1000 reactor in the process of dissolution the simulated spent nuclear fuel in ferric nitrate solutions was studied. Cs, Sr, Ba, Y, La, and Ce together with U pass quantitatively from the fuel into the solution, whereas Mo, Tc, and Ru remain in the resulting insoluble precipitate of basic Fe salt and do not pass into the solution. Nd, Zr, and Pd pass into the solution by approximately 50 %. The recovery of U or jointly U + Pu from the dissolution solution of the oxide nuclear fuel is performed by precipitation of their peroxides, which allows efficient separation of actinides from residues of fission products and iron.     

Comparative evaluation of two substituted diglycolamide extractants for ‘actinide partitioning’

Extraction behaviour of actinides, lanthanides, fission products and structural elements has been studied with the two diglycolamide extractants, namely N,N,N′,N′-tetra-2-ethylhexyl diglycolamide (T2EHDGA) and N,N,N′,N′-tetraoctyl diglycolamide (TODGA). The acid extraction studies suggested that T2EHDGA (K_H: 1.8) is less basic as compared to its linear homologue, TODGA (K_H: 4.1). The distribution ratio of Am(III) by 0.1 M diglycolamides followed the order: TODGA > T2EHDGA. The number of ligand molecules present in the stoichiometry of the extracted species of Am(III) was found to be three and four for T2EHDGA and TODGA, respectively. Thermodynamics studies suggested that the extraction of Am(III) by both the extractants is exothermic in nature. The radiolytic stability of TODGA and T2EHDGA solutions in n-dodecane has been investigated. Due to lower distribution ratio of Am by T2EHDGA, 0.2 M of its solution has been used as compared to 0.1 M solution of TODGA. The distribution behaviour of various metal ions, viz. Am, Nd, Fe, Mo, Cr, Sr and Cs has been studied from nitric acid as well as from simulated high level waste solution.     

New conceptual process for the extraction of actinides based on oxapentane-diamides

Several diamide derivates were synthesized in our laboratory. The extraction of actinides and some fission products by these compounds were studied. N,N,N’,N’-tetra-(2-ethylhexyl)-3-oxa pentanediamide [TEHOPDA) was proven to be a suitable extractant for the removal of actinides from nitric acid solution. The actinides can be stripped from the loaded solvent by the dilute nitric acid. TEHOPDA showed a high loading capacity to actinides and lanthanides with a mixture of n-octanol and kerosene as the diluent. Considering the effective-extraction and easy-stripping of actinides, 0.25 mol/l TEHOPDA — 30% n-octanol + 70% kerosene was selected as the solvent. A cascade extraction experiment was carried out with the simulated dissolver solution of spend fuel as feed. 99.99% U and 99.999% Am, Pu, and Np were extracted in a 4-stage test. Based on the experimental results, a conceptual reprocessing process was proposed.     

Preparation of125I labeled testosterone derivatives for use in radioimmunoassays

A study for separation and sequential recovery of uranium and plutonium from nitric acid solutions by extraction chromatography using tributyl phosphate (TBP)/Amberlite XAD7 as stationary phase is presented. Distribution ratios of actinides, lanthanides and fission products were obtained. The column capacity was investigated and actinides retention conditions were established. Finally, U-Pu sequential separation was studied as well as the U and Pu recovery yields from nitric solutions containing Am/fission products were determined.     

Synthesis of two novel macroporous silica-based impregnated polymeric composites and their application in highly active liquid waste partitioning by extraction chromatography

Two kinds of novel macroporous silica-based chelating polymeric adsorption materials, TODGA/SiO2-P and CMPO/SiO2-P, were synthesized by impregnating and immobilizing two chelating agents, N,N,N',N'-tetraoctyl-3-oxapentane-1,5-diamide (TODGA) and octyl(phenyl)-N,N-diisobutylcarbamoylmethylphoshine oxide (CMPO), into the pores of SiO2-P particles. To separate minor actinides (MA(III)) such as Am(III) and Cm(III), the adsorption and elution of 13 typically simulated fission products from a 3 M HNO3 were performed. It was found that in the first column packed with TODGA/SiO2-P, all of the simulated elements were separated effectively into four groups: (1) Cs(I), Mo(VI), and the most portion of Ru(III) (non-adsorption group), (2) Sr(II), small portion of Gd(III) and all of light REs(III) (MA-lRE-Sr group), (3) most of Gd(III) and all heavy RE(III) (hRE group), and (4) Zr(IV), Pd(II), and a little of Ru(III) (Zr-Pd group) by eluting with 3.0 M HNO3, 1.0M HNO3, distilled water, and 0.5 M H2C2O4, respectively, at 298 K. MA(III) was predicted to flow into the second group along with Nd(III) because of their close adsorption-elution onto TODGA/SiO2-P. In the second column packed with CMPO/SiO2-P, MA-lRE-Sr group was separated into (1) Sr(II), (2) middle RE(III) such as Gd(III), Eu(III), Sm(III), and quite small portion of Nd(III) (MA-mRE), and (3) light RE(III) such as La(III), Ce(III), and most of Nd(III) by eluting with 3.0 M HNO3 and 0.05 M DTPA-pH 2.0, respectively, at 323 K. MA(III) was believed to flow into MA-mRE group along with Gd(III) due to their similar adsorption properties towards CMPO/SiO2-P. Based on positions of MA(III) appeared in light and heavy RE(III), an improved MAREC process for MA(III) partitioning from HLW was proposed.     

Synthesis of N,N’-dimethyl-3-oxa-glutamic acid and application in the simplification of TRPO process

To simplify TRPO process, a novel ligand, N,N’-dimethyl-3-oxa-glutaramic acid (DMOGA), was synthesized and used for stripping of An(III, IV) from 30% TRPO-kerosene. The distribution ratios for transuranium elements, including Np(IV), Pu(IV), Am(III), and some fission products, including Eu(III), Fe and Zr between 30% TRPO-kerosene and various HNO₃-DMOGA solutions were measured. An(III, IV) and Ln(III) extracted to TRPO from simulated high level liquid waste could be recovered with an efficiency of 99.9% above in one stream with a 3-stage crosscurrent strip experiment with 0.2M DMOGA in HNO₃ solution. Using this new agent, the back extraction of TRU elements from loaded TRPO phase becomes more simple and practical. Therefore, the original TRPO process could be simplified.     

Optimizing composition of TODGA/SiO2-P adsorbent for extraction chromatography process

As a part of developing extraction chromatography technology for minor actinides (MA(III); Am and Cm) recovery from spent fast reactor fuels, improvement on the TODGA/SiO2-P adsorbent to enhance its desorption efficiency was carried out. Batchwise adsorption/elution experiments showed that optimizations in amount of the extractant impregnated in the support of the SiO2-P which is the porous silica coated with polymer and degree of the cross linkage of polymer succeeded in finding the optimum values. Inactive column separation experiments with the simulated high level liquid waste and the optimized adsorbent revealed that decontamination factors of fission products can also be improved as well as the recovery yields.     

The EURO-GANEX Process: Current Status of Flowsheet Development and Process Safety Studies

A new hydrometallurgical grouped actinide extraction process has been developed to separate the transuranic actinide ions from dissolved spent fuel solution (after an initial uranium extraction cycle). This “EURO-GANEX” process is aimed towards the homogeneous recycling of plutonium and minor actinides in a future closed fuel cycle. The separation process is based on the co-extraction of actinides and lanthanides from aqueous nitric acid into an organic phase followed by selective co-stripping of actinides. A suitable organic phase has been formulated and distribution ratios determined for lanthanides, actinides and some problematic fission products under extraction and stripping conditions. The process flowsheet has been proven on surrogate feed solutions as well as with spent fast reactor fuel; excellent recoveries of the actinides and good decontamination factors from the lanthanides and other fission products were obtained. A variation on the EURO-GANEX flowsheet (the “TRU-SANEX” process) has now been designed to produce separate Pu+Np and Am+Cm products for heterogeneous recycling. Progress on underpinning process chemistry and safety studies as well as flowsheet tests are summarized.     

Optimization Studies of An(III) and Ln(III) Extraction and Back-Extraction from a PUREX Raffinate by BisDGA Compounds

A family of compounds based on bis-diglycolamide (bisDGA) estructure has been recently developed to be applied for the trivalent actinides and lanthanides (An(III) and Ln(III)) co-extraction by means of DIAMEX process [1], [2]. It has been shown that these bisDGA compounds are efficient extractants of An(III) and Ln(III) regarding to extraction and loading capacity, as well as it has been proved their stability against hydrolysis and radiolysis [3]. For process development, it is necessary to study their selectivity towards An(III) and Ln(III) in the extraction and back-extraction steps in presence of the other elements, such as fission and activation products (FP and AP), in the high active raffinate (HAR) issued from the PUREX process.     

Reversed-phase partition chromatographic separation of minor actinides with bis(2-ethylhexyl) sulfoxide from acidic nitrate PUREX waste solutions

The uptake behavior of U(VI), Pu(IV), Am(III) and a few long-lived fission products from nitric acid media by bis(2-ethylhexyl) sulfoxide (BESO) adsorbed on Chromosorb has been studied U(VI), Pu(IV) and Zr(IV) are taken up appreciably as compared to trivalent actinides/lanthanides including some coexisting fission product contaminants which are weakly sorbed on the column. Chromosorb could be loaded with (1.12±0.03) g of BESO per g of the support. Maximum sorption is observed around 4–5 mol·dm^–3 HNO₃ for both U(VI) and Pu(IV), which are sorbed as their disolvates. The elution of (U(VI) and Pu(IV) from the metal loaded sorbent has also been optimized. Desorption of U(VI) is easily accomplished with dilute nitric acid (ca. 0.01 mol·dm^–3)while Pu(IV) is reductively stripped with 0.1 mol·dm^–3 NH₂OH·HCl. Effective sequential separation of U(VI), Pu(IV) and Am(III) from their several admixtures could be readily achieved from real medium and low level active acidic process raffinates.     

Sequential separation of actinides and lanthanides by extraction chromatography using a CMPO-TBP/XAD7 column

CMPO/TBP sorbed on Amberlite XAD7 resin was used for the separation of actinides and lanthanides from nitric acid solutions by extraction chromatography. The distribution ratios of actinides and lanthanide fission products (Ce, Eu) as a function of acid concentration and some complexing agents were determined. In strong HNO₃ medium (>1 mol/l) the tri-, tetra- and hexavalent actinides as well as the lanthanides have shown great affinity for the CMPO/TBP/XAD7 sorbent. The same behavior was found in HCl medium except for trivalent actinides and lanthanides which show lower distribution values in the same acid range. The effect of some complexing agents as DTPA and ammonium oxalate were also investigated. In DTPA only hexavalent actinides showed higher distribution value. On the basis of these differences, an alternative procedure for actinide-lanthanide separation and actinides from each other is proposed.     

Sorption Capacity of Nitrogen-containing Derivatives of Wood for Ions of Various Metals

Sorption capacity of nitrogen-containing derivatives of aspen wood for Fe(III), Cu(II), Cd(II), and Ca(II) ions was studied. The possibility of using the products formed in oxyammonolysis of wood as sorbents of these metals from dilute aqueous solutions was examined.     

Extraction of actinides with heterocyclic dicarboxamides

Extraction of actinides from aqueous nitric acid by three different heterocyclic dicarboxamides (2,6-pyridinedicarboxamide, 2,2′-bipyridine-6,6′-dicarboxamide and 1,10-phenanthroline-2,9-dicarboxamides) was studied. It was shown that all studied ligands extract actinides at different oxidation states (U(VI), Np(V), Pu(IV), Am(III), Cm(III)) from acidic solutions. All studied diamides extract Am(III) better than Cm(III). Et(pHexPh)ClPhen contains electron-withdrawing chlorine atoms at the positions 4 and 7 of the phenanthroline moiety (SF_Am/Cm = 4–6) and possesses the highest separation factor Am(III)/Cm(III). The studied ligands possess high extraction ability to all actinides present in HLW and therefore they could be used for simultaneous extraction of actinides in the GANEX-type process.     

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.     

A pulse radiolysis investigation of the reactions of tributyl phosphate with the radical products of aqueous nitric acid irradiation

Tributyl phosphate (TBP) is the most common organic compound used in liquid-liquid separations for the recovery of uranium, neptunium, and plutonium from acidic nuclear fuel dissolutions. The goal of these processes is to extract the actinides while leaving fission products in the acidic, aqueous phase. However, the radiolytic degradation of TBP has been shown to reduce separation factors of the actinides from fission products and to impede the back-extraction of the actinides during stripping. As most previous investigations of the radiation chemistry of TBP have focused on steady state radiolysis and stable product identification, with dibutylphosphoric acid (HDBP) invariably being the major product, here we have determined room temperature rate constants for the reactions of TBP and HDBP with the hydroxyl radical [(5.00 +/- 0.05) x 10(9), (4.40 +/- 0.13) x 10(9) M(-1) s(-1)], hydrogen atom [(1.8 +/-0.2) x 10(8), (1.1 +/- 0.1) x 10(8) M(-1) s(-1)], nitrate radical [(4.3 +/- 0.7) x 10(6), (2.9 +/- 0.2) x 10(6) M(-1) s(-1)], and nitrite radical (<2 x 10 (5), <2 x 10(5) M(-1) s(-1)), respectively. These data are used to discuss the mechanism of TBP radical-induced degradation.     

Improving the uptake capacity of natural clays by phosphate and heat treatment

Phosphate treatment of carbonaceous clays increased the uptake capacity for uranium and thorium. Heat treatment up to 400°C improved enormously the uptake property of the treated clay. These treatments failed to improve the uptake quality of argileous clays. Strontium exchanges weakly with clays of high Ca and Mg contents but strongly with a silicoaluminium clay of low Ca and Mg contents. A mixed bed of these two clays was found adequate for decontamination of actinides and fission products.     

Hydrous oxides in the concentration and removal of fission products from alkaline solutions of uranium

A great variety in retention properties occurs as a result of different methods of preparation of the sorbents. Specific surfaces and porosities, which are mostly connected with the sorption activities, may vary widely. The activation of Al₂O₃, SnO₂ and silica gel, to produce highly active sorbents occurs only if the oxide is contacted with acid immediately after thermal treatment. The efficiency of the separation scheme has been tested using uranium and fission products under static conditions from strongly alkaline aqueous solutions.     

Evaluation of tertiary pyridine resin for the separation of lanthanides by simulated moving‐bed chromatography

Lanthanide separation by simulated moving‐bed chromatography was studied as a model system for separating lanthanide fission products and minor actinides from used nuclear fuels. The simulated moving‐bed system was modeled for a tertiary pyridine anion‐exchange resin supported on silica particles as the stationary phase and a mixture of methanol and 1M nitric acid as the mobile phase. Pulse injection tests using a single packed column were used to obtain chromatographic parameters for mathematical modeling of the simulated moving‐bed system. Higher concentrations of methanol improved the separation, but the chromatograms showed evidence of nonlinearity of the isotherms. The mathematical model of the simulated moving‐bed process predicted a production rate of purified samarium and neodymium at 118 g solute/L resin/day and a purity of 99.5%. The optimal methanol ratio for the production rate for various product purities was determined from the model. The excellent separation of Nd and Sm suggests that the simulated moving‐bed system could be applied to the separation of minor actinides such as americium and curium.     

Efficient CO2 capture on low-cost silica gel modified by polyethyleneimine

In this work,a series of polyethyleneimine(PEI) functionalized commercial silica gel were prepared by wet impregnation method and used as CO 2 sorbent.The as-prepared sorbents were characterized by N 2 adsorption,FT-IR and SEM techniques.CO 2 capture was tested in a fixed bed reactor using a simulated flue gas containing 15.1% CO 2 in a temperature range of 25-100 C.The effects of sorption temperature and amine content on CO 2 uptake of the adsorbents were investigated.The silica gel with a 30 wt% PEI loading manifested the largest CO 2 uptake of 93.4 mg CO 2 /g adsorbent(equal to 311.3 mg CO 2 /g PEI) among the tested sorbents under the conditions of 15.1%(v/v) CO 2 in N 2 at 75 C and atmospheric pressure.Moreover,it was rather low-cost.In addition,the PEI-impregnated silica gel exhibited stable adsorption-desorption behavior during 5 consecutive test cycles.These results suggest that the PEI-impregnated silica gel is a promising and cost-effective sorbent for CO 2 capture from flue gas and other stationary sources with low CO 2 concentration.