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.     

Modeling of Pu(VI) distribution coefficients in 30 % TBP based PUREX process

Hexavalent plutonium (Pu(VI)) is an important solute in the PUREX (plutonium uranium extraction) process. In 30 % TBP based PUREX solvent extraction system, distribution coefficient of Pu(VI) is much lower than that of Pu(IV). This lower distribution coefficient of Pu(VI) may cause unexpected Pu loss during primary HA extraction in low acid flowsheets. An empirical model for Pu(VI) distribution coefficients in 30 % TBP and its temperature dependency has been reported in this paper. Comparison with literature data revealed a reasonably good agreement between the reported experimental and model predicted values.     

CARBEX process,a new technology of reprocessing of spent nuclear fuel

Advances in the CARBEX process, a new aqueous chemical method for reprocessing of spent nuclear fuel (SNF) in carbonate media, are considered. A review of carbonate methods for SNF reprocessing is given. The CARBEX process concept is presented and experimental data for every stage of the CARBEX process: high-temperature oxidation of spent fuel composition, its oxidative dissolution in carbonate aqueous solutions, extraction refining of U(VI) and Pu(VI), solid-phase re-extraction of carbonate complexes of U(VI) and Pu(VI), and obtaining of uranium and plutonium dioxide powders for fabrication of ceramic nuclear fuel, are discussed. It was shown that the CARBEX process can be more effective and safe than the well-known industrial PUREX process.     

Tri-iso-amyl phosphate (TAP): An alternative extractant to tri-butyl phosphate (TBP) for reactor fuel reprocessing

Tri-iso-amyl phosphate (TAP), an indigenously prepared extractant was utilized for reactor fuel reprocessing and compared with tri-butyl phosphate (TBP) and tri-n-hexyl phosphate (THP). The potential of these extractants was found to be in the order TAP>THP>TBP by calculating the acid uptake value (K _H). The effect of various parameters such as solvent degradation due to acid hydrolysis, radiation effect, decontamination factor and phase separation were investigated and it was found that TAP was always a better extractant in comparison to THP and TBP. In addition to this, the extraction of fission product contaminants such as ¹⁴⁴Ce, ¹³⁷Cs, ¹⁰⁶Ru, ⁹⁵Zr was almost negligible, even at very high nitric acid concentrations in the aqueous phase, indicating the potential application of TAP in actinide partitioning. Sodium carbonate solution or acidified distilled water was a good strippant for U(VI), similarly, uranium(IV) nitrate stripped Pu(IV) from the organic phase.     

Hot Test in Mixer Settlers of Alpha Barrier with AHA in PUREX Process

In the PUREX process, the first U-Pu purification cycle (1CUPu) is not efficient enough for the decontamination of uranium flow out of neptunium. In this context, molecules known for their strong complexing power for actinides(IV) in aqueous phase, such as acetohydroxamic acid (AHA) have been tested in batch experiments to strip Np and Pu from TBP solvent loaded with U. A phenomenological model was developed and with the help of this model, a flowsheet of a counter-current alpha barrier process was designed and tested in C17 glove boxes in ATALANTE facility. A decontamination factor DFU/Np of 480 was obtained, higher than DFU/Np required by UNIREP standards.     

Technetium chemistry in a novel group actinide extraction process

A newly developed method for advanced reprocessing of used nuclear fuel is the Group ActiNide EXtraction (GANEX) process. It is a liquid–liquid extraction process that aims at extracting all the actinides as a group from dissolved used nuclear fuel. This extraction can either be performed after a removal of the bulk uranium or directly on the dissolution liquor. At Chalmers University of Technology in Sweden a solvent that utilizes tributyl-phosphate (TBP) and a molecule from the bis-triazine bipyridine (BTBP) class of ligands dissolved in cyclohexanone has been developed for the use in a GANEX process. Previously the system has not been tested with the presence of technetium that is one of the major fission products. Technetium is often considered a problem within reprocessing since it has a chemical behaviour that differs from most other elements in the spent fuel. Therefore, a special emphasis was put on the investigation of technetium in the selected GANEX system. It was shown that technetium is readily extracted by the GANEX solvent and that cyclohexanone is the main extractant when no other metals were present in the system. It was also found that the presence of uranium decreased the overall technetium extraction despite a slight co-extraction with TBP, while irradiation of the GANEX solvent to large doses (>1 MGy) increased its technetium extraction capability. It was also discovered that an increased nitrate concentration in the aqueous phase and an addition of other fission products both inhibited the technetium extraction even though fission product loading most likely changed the extraction mechanism to co-extraction by BTBP.     

Synthesis and characterization of red-oil from tri <Emphasis Type="Italic">iso</Emphasis>-amyl phosphate/<Emphasis Type="Italic">n</Emphasis>-dodecane/nitric acid mixtures at elevated temperature

Red-oil is a mixture of nonspecific composition consisting of extractant, degradation products, nitrated solvent and unidentified red-coloured nitro-organics. Red-oil formation is coupled with decomposition of extractant and diluent into gases of explosive nature. If ignited or incinerated, these gases may cause rapid pressurization and endanger the integrity of containment. Such an event occurred at Tomsk-7 facility in 1993. To ensure safe operation, red-oil formation has to be avoided in the fuel cycle facilities by a careful combination of several independent measures like strict control over temperature, limiting organic entrainment in the aqueous streams (which are to be concentrated by evaporation) and control over acidity of aqueous phases. Since tri-iso amyl phosphate (TiAP) has much lower aqueous solubility as compared to TBP, it is visualized as alternate solvent for PUREX process. In this work, TiAP red-oil was synthesized and characterized.     

Sequential separation of transuranic elements and fission products from uranium metal ingots in electrolytic reduction process of spent PWR fuels

A sequential separation procedure has been developed for the determination of transuranic elements and fission products in uranium metal ingot samples from an electrolytic reduction process for a metallization of uranium dioxide to uranium metal in a medium of LiCl-Li₂O molten salt at 650 °C. Pu, Np and U were separated using anion-exchange and tri-n-butylphosphate (TBP) extraction chromatography. Cs, Sr, Ba, Ce, Pr, Nd, Sm, Eu, Gd, Zr and Mo were separated in several groups from Am and Cm using TBP and di(2-ethylhexyl)phosphoric acid (HDEHP) extraction chromatography. Effect of Fe, Ni, Cr and Mg, which were corrosion products formed through the process, on the separation of the analytes was investigated in detail. The validity of the separation procedure was evaluated by measuring the recovery of the stable metals and ²³⁹Pu, ²³⁷Np, ²⁴¹Am and ²⁴⁴Cm added to a synthetic uranium metal ingot dissolved solution.     

Evaluation of long chain monoamide extractants for the reprocessing of U–Pu–Zr metallic fuel solution

Long chain monoamide extractants, N,N-di-decyloctanamide(DDOA), N,N-di-hexyldecanamide(DHDA), N,N-di-2-ethylhexyloctanamide(D2EHOA) and N,N-dihexyl-2-ethylhexanamide(DH2EHA) were synthesized and studied for the recovery of U(VI), Pu(IV) and Zr(IV) from a simulated dissolver solution of un-irradiated U–Zr metallic fuel. The results were compared with the results of N,N-dihexyloctanamide(DHOA) and tri-n-butylphosphate(TBP) under similar conditions. Solvent extraction studies were carried out for comparing the extraction behavior of U(VI), Pu(IV) and Zr(IV) in monoamide extractants with TBP system. The influence of length and branching of alkyl chains on either side of the amidic group on the extraction efficiency, third phase behaviour and metal ion selectivity in long chain monoamides has been discussed based on the results of above studies.

     

Chemical separation and ICP-AES determination of 22 metallic elements in U and Pu matrices using cyanex-923 extractant and studies on stripping of U and Pu

Comprehensive studies have been carried out on the extraction behavior of uranium and plutonium matrices using cyanex-923 extractant. The near total extraction of U/Pu and quantitative separation of 22 metallic elements at trace levels has been established using inductively coupled plasma-atomic emission spectrometry (ICP-AES). The studies carried out on back extraction of U/Pu from organic phase have established the near total recovery of these matrices into the aqueous phase using 1 M Na(2)CO(3) and saturated oxalic acid, respectively.     

N,N-Dialkyl amides as extractants for spent fuel reprocessing: an overview

Reprocessing of spent nuclear fuel is vital for the long-term global nuclear power growth and is the major motivation for developing novel separation schemes. Conventionally, PUREX and THOREX processes have been proposed for the reprocessing of U and Th based spent fuels employing tri-n-butyl phosphate (TBP) as extractant. However, based on the experiences gained over last five–six decades on the reprocessing of spent fuels, some major drawbacks of TBP have been identified. Evaluation of alternative extractants is, therefore, desirable which can overcome at least some of these problems. Extensive studies have been carried out on the evaluation of N,N-dialkyl amides as extractants in the back-end of the nuclear fuel cycle for addressing the issues related to the reprocessing of U and Th based spent fuels. Under advanced fuel cycle scenario, efforts are also being made by countries with a developed nuclear technological base to provide safe nuclear power to other countries and to minimize proliferation concerns worldwide. This paper presents an overview of studies carried out in our laboratory on different aspects of reprocessing of U and Th based spent fuels employing N,N-dialkyl amides as extractants.     

Radiation chemistry of cis-syn-cis dicyclohexano-18-crown-6 (DCH18C6): Acidity and uranyl nitrate dependence

The cis-syn-cis isomer of dicyclohexano-18-crown-6 (DCH18C6) has been shown to be an efficient extractant able to perform the separation of Pu(IV) and U(VI) from fission products and then the separation of Pu(IV) from U(VI) without valence exchange as required in the PUREX process. This macrocycle was irradiated in nitric acid with a ¹³⁷Cs γ source to study its radiation chemical stability. Radiation chemical yields (G) were determined by gas chromatography. The results show that the presence of uranyl nitrate has a strong influence on DCH18C6 radiation chemical stability. Indeed, the presence of this template ion increases the macrocycle stability by promoting fragments recombination.     

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.     

Improved determination of tributyl phosphate degradation products (mono- and dibutyl phosphates) by ion chromatography

Tributyl phosphate (TBP) is a very important compound in the nuclear industry, particularly in the area of nuclear fuel reprocessing. This compound is used in the PUREX (plutonium and uranium refining extraction) process which consists of the extraction of uranium and plutonium from an aqueous nitric acid phase, for the purpose of recycling. But TBP may be degraded to dibutyl phosphate (DBP) and monobutyl phosphate (MBP) by dealkylation of one or two butoxy groups, respectively. We have compared and evaluated the capacity of two resins manufactured by Dionex (AS11 and AS5A) in the separation and measurement of these two degradation products. AS11 generates two interferences: nitrite/DBP and carbonate/MBP. The first one is the most serious. So, we have developed a method for oxidising nitrite ions to nitrate ions which have no trouble over the measurement. The second resin tested, AS5A, allows a very efficient separation between DBP and NO₂⁻ ions and a good separation between MBP and CO₃²⁻ in comparison with the AS11. The detection limits for the AS5A column are 0.13 μM for MBP and 0.71 μM for DBP (injection LOOP=50 μl).     

An improved PUREX process in technetium separation by stepwise reduction

An improved PUREX process in Tc separation is introduced in this paper. Experiments including we did and done by other investigators are cited in this paper to testify the feasibility of the process. The scheme of this process is as follows: First, to extract U, Pu and most Tc, Np into the organic phase, so the concentration of Tc in the high level liquid waste (HLLW) may be very low, which can alleviate the burden of Tc treatment in HLLW vitrification. Second, in the Pu and Np separation stage to reduce or complex Pu and Np using acetohydroxamic acid, this step separates Pu and Np from the organic phase. Next, to separate Tc from U, it can be realized easily by reducing Tc to lower valance using reductant such as hydrazine or quadrivalence U. By this technology, we may resolve the problems caused by Tc remained in the HLLW such as a long time surveillance of HLLW condensate or the risk of Tc transferring into the biosphere, and meanwhile the over-consumption of reductant in the U/Pu splitting stage can be avoided.     

Studies on sorption of plutonium from carbonate medium on polyacrylhydroxamic acid resin

Polyacrylhydroxamic acid resin synthesized by functionalization of polyacrylamide with hydroxylamine has been investigated for the sorption of plutonium(IV) from carbonate medium, aiming at its application for the removal of plutonium from alkali wash effluent generated during purification of TBP in PUREX process. Batch experiments have been carried out to determine distribution coefficient of plutonium(IV) between this exchanger and various compositions of carbonate medium. Effect of the concentration of sodium carbonate, sodium bicarbonate and pH of the solution on the distribution coefficient have been studied to optimize the conditions for the uptake of Pu(IV) by this exchanger. Column experiments were carried out to determine the practical capacity of the exchanger for plutonium. Elution studies were also carried out to recover the loaded plutonium from the ion exchange column The exchanger displayed good exchange capacity for Pu(IV) from feed solution simulating the conditions of carbonate wash effluent generated in PUREX process. The exchanger also exhibited fast elution of Pu, suggesting the feasibility of using it for the recovery of Pu from carbonate based wash effluent.     

Synthesis of dihydroxyurea and its application to the U/Pu split in the PUREX process

Dihydroxyurea (DHU) was synthesized using tri-associated solid phosgene [bis(trichloromethyl) carbonate] dissolved in dioxane and hydroxylamine hydrochloride dissolved in potassium acetate solution. The reduction of Pu(IV) by DHU was investigated using UV-Vis spectrophotometry. The reduction back-extraction behavior of Pu(IV) in 30% tri-butyl phosphate/kerosene was firstly investigated under conditions of various temperature, various DHU and HNO₃ concentrations and various phase contact times. The results showed that Pu(IV) in the organic phase can be stripped rapidly to the aqueous phase by DHU. Simulating the 1B contactor of the PUREX process using a 0.1 M DHU in 0.36M nitric acid solution as the stripping agent, the separation factors of uranium/plutonium can reach 2.1·10⁴. This indicates that DHU is a promising salt free agent for uranium/plutonium separation.     

Separation and Recovery of Uranium and Plutonium from Oxalate Supernatant

The separation of uranium and plutonium from oxalate supernatant, obtained after precipitating plutonium oxalate, containing ~10 g/l uranium and 30–100 mg/l plutonium in 3M HNO₃ and 0.10–0.18M oxalic acid solution has been carried out. In one extraction step with 30% TBP in dodecane: ~92% of uranium and ~7% of Pu is extracted. The raffinate containing the remaining U and Pu is extracted with 0.2M CMPO+1.2 M TBP in dodecane and near complete extraction of both the metal ions is achieved. The metal ions are back extracted from organic phases using suitable stripping agents. The recovery of both the metal ions separately is >99%. The uranium species extracted into the TBP phase from the HNO₃+oxalic acid medium was identified as UO₂(NO₃)₂·2TBP.     

Extraktionschromatographische Trennung von Uran,Plutonium und Spaltprodukten

An extraction chromatographic method has been developed to separate plutonium and uranium from irradiated uranium fuel samples. Tributylphosphate fixed on polytetrafluoro ethylene was used as stationary phase and HNO₃ of various concentrations as mobile phase. The separation of plutonium by reduction with H₂SO₃ was studied in column experiments. The method was applied to the separation of irradiated U/Mg-fuel with a Pu/U ratio of 1∶400.     

The effect of previous gamma-irradiation on the extraction of U(VI), Th(IV), Zr(IV), Eu(III) and Am(III) by various amides

The extraction behavior of U(VI), Th(IV), Zr(IV), Eu(III) and Am(III) from 3.5M nitric acid with a series of gamma-pre-irradiated symmetrical and unsymmetrical monoamides in benzene has been investigated up to a dose of 100 Mrad. The results indicated that the radiolytic stability is influenced by the structure of amides. Symmetrical monoamides seem to be less affected by radiation compared with unsymmetrical monoamides. Infrared studies identify the final products of radiolysis as the respective carboxylic acids and amines. The radiolytic degradation of the investigated monoamides has been estimated by quantitative IR spectroscopy. Extraction data obtained under similar experimental conditions for U(VI), Th(IV) and Zr(IV) with the TBP/benzene system have also been compared. This revised version was published online in August 2006 with corrections to the Cover Date.