Purex Process Research Articles

Study on the technology of U-Pu Co-stripping process for reprocessing spent fuel of fast reactor

Abstract To develop an aqueous reprocessing process for spent fuel of fast reactor, a counter current experiment (U-Pu co-stripping) based on a simplified PUREX process was carried out with 30%TBP/Kerosene solution containing U, Pu and HNO3. The experiment was done in a 16-stage miniature centrifugal contactor. The stripping percentage of U was 99.99%, Pu 99.94%. U stripping mainly took place in X1-feeding stage (11.00 mol/L HNO3 feeding stage) to the 16th one, while Pu stripping mainly took place in the first stage to the X1-feeding stage. And a simple method was presented to calculate c U/X1−feeding stage, Org when the X1-feeding stage number is more than 7 and the stripping percentage of U from the X1-feeding stage is more than 99.9% through the rear stages' stripping. The calculation results by this method were validated by Computer Simulation when only changing U concentrations in feed solution from 40.0 g/L to 90.0 g/L or only changing the X1-feeding stage from the 7th stage to the 11th one. Finally the optimized technology was given.

Theoretical study of plutonium(IV) complexes formed within the PUREX process: a proposal of a plutonium surrogate in fire conditions.

We present a relativistic quantum chemical study to determine the best surrogate for plutonium(IV) to be used in experimental investigations of the behavior of plutonium-nitrate-TBP in fire conditions that might occur in the nuclear fuel refining process known as PUREX. In this study geometries and stabilities of Pu(NO3)6(2-) and Pu(NO3)4(TBP)2 complexes were compared to that of equivalent complexes of selected elements from the lanthanide and actinide series (Ce, Th, U) chosen on the basis of similar ionic radii and stability as tetravalent species. PBE and PBE0 DFT functionals have proven to be sufficient and affordable for qualitative studies, performing as good as the wave function based correlated method MP2. On the basis of our results, cerium(IV) appears to be a good surrogate for plutonium(IV).

Ways of technetium and neptunium localization in extraction reprocessing of spent nuclear fuel from nuclear power plants

Possible ways of Tc and Np localization within the framework of the first extraction cycle of the Purex process are considered. Experimental data are presented on Tc localization in the high-level raffinate as applied to the RT-1 flowsheet, in the zirconium-tritium strip, and in the Pu-Np product or in the scrub from the unit of the barrier reductive scrubbing of the uranium solvent product as applied to six-unit extraction scheme in the design of the Experimental and Demonstration Center. For the latter case, the possibility of Np localization before and after Pu stripping was also examined.

Effects of temperature, concentration of acid and metal ions on the solubility of tri-n-butyl phosphate in aqueous phase

The solubility of Tri-n-butyl phosphate (TBP) in nitric acid (0–14 M) was measured in the temperature range 25–60 °C for the purpose of understanding the distribution equilibrium of TBP between aqueous and organic phases as well as to provide input for designing a suitable contactor in order to remove the dissolved TBP from aqueous streams in the PUREX process. With increase in temperature, solubility of TBP was observed to increase and with increase in acid concentration, the solubility exhibited a minimum and had shown an increasing trend above 10 M acid. The effect of zirconium, ruthenium and uranium on the solubility of TBP in aqueous phase was also estimated.

Study of the plutonium (IV) electrochemical behavior in nitric acid at a platinum electrode. Application to the cathodic reduction of Pu(IV) in a plate electrolyzer

Spent fuel reprocessing and notably plutonium recycling contributes in enhancing the sustainability of nuclear energy by preserving uranium natural resources and decreasing the long term radiotoxicity of the final waste. After initial dissolution of the spent nuclear fuel in nitric acid, uranium and plutonium recovery from ultimate waste (fission products and minor actinides) is operated by a liquid–liquid extraction process (PUREX process) and plutonium partitioning from uranium is run on the basis of the selective reduction of Pu and its back extraction. Therefore, the redox properties of Pu in nitric acid are of prime importance and the stabilization of Pu at its lower redox state+III is to be mastered. This study then deals with the characterization of Pu(IV)/Pu(III) redox couple in aqueous nitric acid solution. It was aimed at measuring physico-chemical features of the redox couple such as half-wave and conditional standard potentials, diffusion coefficients and charge transfer rate constant in nitric acid medium. These features have been used in modeling experimental results of preparative Pu(IV) reduction tests in nitrate medium in a laboratory parallel plate electrode cell.

Determination of dissolved TBP in PUREX process aqueous streams using ICP-AES

A method for the determination of dissolved tri butyl phosphate (TBP) in PUREX process aqueous streams by atomic emission spectrometry with inductively coupled plasma has been evolved. The phosphorous emission lines of 213.618 and 214.914 nm have been chosen and the possibility of determining dissolved TBP in these lines has been explored. Since presence of uranium which is inevitable in the process solution, and also present in macro level, the spectral and matrix interference of uranium on the determination of phosphorous has been investigated. The precision of the method was characterized by RSD of <5 % with a detection limit of 0.06 and 0.24 mg/L for phosphorous which in turn corresponds to detection limit of 0.5 and 2 mg/L for dissolved TBP in lines 213.618 and 214.914 nm respectively.

Chemical behavior of neptunium in the presence of technetium in nitric acid media

Abstract The chemical behavior of Np is influenced by many factors, such as U(IV), hydrazine, Pu and Tc, which coexist in U/Pu separation stage of PUREX process. Firstly, the reduction of Np(V) to Np(IV) by Tc(IV) was studied in nitric acid media, and the rate equation could be expressed as − d[Np(V)]/dt = k[Np(V)][Tc(IV)]0.8[HNO3]1.2, k = 28.5 ± 0.9 (L/mol)2/min at 25 ℃, the activation energy was E a = 70.0 kJ/mol. Thereafter, the chemical behavior of Np was investigated in aqueous nitric acid solution containing Tc(VII), U(IV) and N2H4. Np(V) was reduced to Np(IV) quickly, Tc(VII) was reduced to Tc(IV) and U(IV) was oxidized to U(VI) at the same time. Thereafter Np(IV) was entirely reoxidized to Np(V) when U(IV) and hydrazine were completely consumed. The fast reduction of Np(V) was mainly ascribed to Tc(IV). In addition, Tc(VII) and nitric acid had little influence on the fast oxidation of Np(IV).

An advanced purex process based on salt-free reductants

Abstract An advanced plutonium and uranium recovery process has been established based on two organic reductants, N,N-dimethylhydroxylamine (DMHAN) and methylhydrazine (MH), as U/Pu separation reagents. This Advanced Purex process based on Organic Reductants (APOR) is composed of three cycles, including U/Pu co-decontamination/separation cycle, uranium purification cycle and plutonium purification cycle. Using DMHAN and MH as plutonium stripping reagents in the U/Pu co-decontamination/separation cycle and plutonium purification cycle, the APOR process exhibits high performance with following highlights: 1) the process is much simpler because of the elimination of Tc scrubbing operation and the supplement extraction operation, 2) high efficiency of U/Pu separation can be achieved in the first cycle, 3) plutonium product solution of high concentration can be obtained in the Pu purification cycle with a simple extraction operation instead of circumfluent extraction or evaporation of the plutonium solution.

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.

Electrolytic partitioning of uranium and plutonium based on a new type of electrolytic mixer-settler

Summary The design of a new type of electroreduction mixer-settler for the partitioning of uranium and plutonium during the Purex process, which is featured with E-shaped cathodes and U-shaped anodes in settling chamber, is described and the operational results achieved using this equipment are presented. The results show that this new type of mixer-settler has excellent separation performances. The flow rate of organic feed solution is 3 mL/min and the flow ratio of feed solution (1BF) to aqueous back extraction stream (1BX) and to organic wash stream (1BS) is 4/1/1. For an organic feed of 84 g/L uranium and 1.40 ⁓ 2.64 g/L plutonium, both the separation factor of plutonium from uranium and that of uranium from plutonium are apparently higher than 104

Predominance diagrams of uranium and plutonum species in both lithium chloride–potassium chloride eutectic and calcium chloride

Electro-reduction of spent nuclear fuel has the potential to significantly reduce the amount of high level waste from nuclear reactors. Typically, spent uranium and plutonium are recovered via the PUREX process leading to a weapons-grade recovery; however, electro-reduction would allow spent nuclear fuel to be recovered effectively whilst maintaining proliferation resistance. Here, we present predominance diagrams (also known as Littlewood diagrams) for both uranium and plutonium species in molten lithium chloride–potassium chloride eutectic (LKE) at 500 °C and in calcium chloride at 800 °C. All diagrams presented depict regions of stability of various phases at unit activity in equilibrium with their respective dissociated ions. The diagrams thermodynamically define the electrochemical system leading to predictions of reaction conditions necessary to electrochemically separate species. The diagrams have been constructed using a pure thermodynamic route; identifying stable species within the molten salt with an assumption of unit activity for each of the phases. These thermodynamically predicted diagrams have been compared to the limited available experimental data; demonstrating good correlation. The diagrams can also be used to predict regions of stability at activities less than unity and is also demonstrated.

Assessment of proliferation resistances of aqueous reprocessing techniques using the TOPS methodology

The aim of this study is to assess and compare the proliferation resistances (PR) of three possible Generation IV lead-cooled fast reactor fuel cycles, involving the reprocessing techniques Purex, Ganex and a combination of Purex, Diamex and Sanex, respectively. The examined fuel cycle stages are reactor operation, reprocessing and fuel fabrication. The TOPS methodology has been chosen for the PR assessment, and the only threat studied is the case where a technically advanced state diverts nuclear material covertly.According to the TOPS methodology, the facilities have been divided into segments, here roughly representing the different forms of nuclear material occurring in each examined fuel cycle stage. For each segment, various proliferation barriers have been assessed. The results make it possible to pinpoint where the facilities can be improved.The results show that the proliferation resistance of a fuel cycle involving recycling of minor actinides is higher than for the traditional Purex reprocessing cycle. Furthermore, for the purpose of nuclear safeguards, group actinide extraction should be preferred over reprocessing options where pure plutonium streams occur. This is due to the fact that a solution containing minor actinides is less attractive to a proliferator than a pure Pu solution. Thus, the safeguards analysis speaks in favor of Ganex as opposed to the Purex process.

Separation and purification of 106Ru from effluent streams of ion-exchange cycle used for Pu purification in PUREX process

This paper describes the separation and purification of 106Ru from ion-exchange effluent streams of PUREX process using combinations of solvent extraction, extraction chromatography, and ion-exchange. The effluent stream from ion-exchange cycle contains ∼100 mCi/L 106Ru activity along with contaminants like 95Nb, uranium, and plutonium at an acidity of ∼7.00M HNO3. In first step, the feed solution is contacted with 30% Tri-n-butyl phosphate in n-dodecane which not only lowers the acidity of the feed solution but also reduces the Pu, U, and Nb concentration significantly. The raffinate from this step is subjected to extraction chromatography wherein uranium and plutonium are completely sorbed on a column containing 2-ethylhexyl 2-ethylhexyl phosphonic acid (KSM-17) sorbed on an inert polymeric support XAD-4. In these steps, Ru activity is found to retain in the aqueous raffinate/effluent phase without any loss. The effluent from the second step is neutralized to an acidity of ∼0.50M using NaOH solution and passed through a silica gel column for complete sorption of Nb activity. The effluent obtained from the silica gel column shows 106Ru, free from all the radioactive contaminants. The recovery of Ru separation is found to be ##xgt;90%. Based on results, a flow sheet is proposed and used for Ru separation at mCi levels.

Separation and recovery of ruthenium: a review

The chemistry of the noble metal fission product, ruthenium is very complex due to the existence of many oxidation states in addition to forming a large number of co-ordination complexes. In the PUREX process for the separation of U and Pu from the spent nuclear fuels from fast breeder reactors, owing to the high volatile nature of RuO4 problems arise not only during the extraction stages but also in the treatment of high active liquid waste and subsequent vitrification. As this volatile RuO4 can deposit in cooler parts, there is an increase in the radiation field due to the presence of 106Ru. The problem is very acute in the reprocessing of fast reactor fuels due to the increased concentration of ruthenium in the spent fuel. In nitric acid medium Ru can exist in various nitroso nitrate complexes and nitroso complexes are more stable than nitrates. The nitrates are non-extractable by the solvent TBP; however, they are extractable to a higher degree by DBP (the primary degradation product of TBP). The extractability of Ru nitrates into the solvent is inhibited by high acid content, temperature and prolonged hold-up time. Nevertheless, these factors promote the volatilization of Ru as RuO4. The volatilization is enhanced by the addition of phosphate ions, but is suppressed by phosphite or hypophosphite ions. Thus, it would be advantageous if ruthenium is removed so that not only the purity of the product (Pu) is improved, but also the problem related to volatilisation can be resolved. High molecular weight amines (tertiary amines) capable of forming co-ordinate bonds are reported to be ideal extractants for Ru. Gas phase separation is an effective method for the recovery of Ru from catalysts, lead button and from other platinum group metals. Separation and pre-concentration of noble metals can be accomplished from non-metals by simple sorbents like coconut shell activated carbon to complicated chelating resins, aromatic polymers and zeolites. In the electro-oxidation of active Ru from nitroso salts, Pd was found to interfere and removal of Pd prior to oxidation of Ru is recommended. Redox catalysts such as Ag2+ and Ce4+ are found to play a prominent role in the electro-oxidation of Ru. Though, various methods and extractants are reported in the literature for the separation of Ru, R&D is being pursued for the removal of Ru during aqueous reprocessing of spent fuels using extractants and methods which are conducive to plant conditions. Hence, an exhaustive survey of literature was made and the different methods reported for the removal of Ru with emphasis towards reprocessing applications are discussed in this report as a review. Attempts made by the authors in separating Ru from simulated waste solution are also included in this review.

Estimation of 90Sr in reprocessing streams using pre-packed extraction chromatographic column of 4,4'(5')-bis(tert-butylcyclohexano)-18-crown-6 impregnated on amberlite XAD-7

This paper describes a method for the estimation of 90Sr in various PUREX and THOREX process streams. Radiostrontium present in the sample is loaded on an extraction chromatographic column containing 4,4’(5’)-bis(tert-butylcyclohexano)-18-crown-6 (DtBuCH18C6) in 1-octanol impregnated on amberlite XAD-7 after diluting it appropriately in about 3.0 M HNO3. The column is then washed using 3 M HNO3 to remove the impurities completely and finally 90Sr is eluted in 5 mL of water and subjected to beta counting without time delay. Recovery of 90Sr separation is studied in presence of varying quantity of uranium and thorium in the sample. The method is used to estimate 90Sr in actual process samples containing 90Sr activity in wide range of activity varying from Bq/mL to Ci/L. The results are compared with those obtained by conventional radiochemical precipitation method and found to agree within ±5 %.

Studies on Transport of Anionic Complex of Plutonium From Nitric Acid Medium across ALIQUAT 336/n-Paraffin Liquid Membrane

The present studies deal with the application of the supported liquid membrane (SLM) technique for the separation and purification of plutonium from other impurities in nitric acid medium using anion exchanger Aliquat 336 (a quaternary ammonium salt)/ n-paraffin as a carrier. The effects of feed acidity, stripping agent, and membrane pore size and membrane thickness on the transport behavior of anionic complex of plutonium have been studied in detail. An attempt has been made to establish the mechanism for plutonium transport and model the physicochemical transport of plutonium across SLM. Transport of anionic complex of plutonium increased with increase in carrier concentration upto 10% (w/v), while with further increase in carrier concentration, decrease in transport of plutonium was observed. The PTFE membrane with 0.45 µm pore size and 80 µm thickness was found to be most suitable for the transport of plutonium. The effect of membrane thickness indicates that the transport phenomenon is diffusion controlled. Transport behavior of plutonium, uranium, and other fission products from actual feed solution of ion exchange method obtained in PUREX process was also tested and the result clearly indicates that Aliquat 336 has high selectivity for plutonium and it can be used for the separation and purification of plutonium by the supported liquid membrane technique.

Comparison of actinides and fission products recycling scheme with the normal plutonium recycling scheme in fast reactors

Multiple recycling of actinides and non-volatile fission products in fast reactors through the dry re-fabrication/reprocessing atomics international reduction oxidation process has been studied as a possible way to reduce the long-term potential hazard of nuclear waste compared to that resulting from reprocessing in a wet PUREX process. Calculations have been made to compare the actinides and fission products recycling scheme with the normal plutonium recycling scheme in a fast reactor. For this purpose, the Karlsruhe version of isotope generation and depletion code, KORIGEN, has been modified accordingly. An entirely novel fission product yields library for fast reactors has been created which has replaced the old KORIGEN fission products library. For the purposes of this study, the standard 26 groups data set, KFKINR, developed at Forschungszentrum Karlsruhe, Germany, has been extended by the addition of the cross-sections of 13 important actinides and 68 most important fission products. It has been confirmed that these 68 fission products constitute about 95% of the total fission products yield and about 99.5% of the total absorption due to fission products in fast reactors. The amount of fissile material required to guarantee the criticality of the reactor during recycling schemes has also been investigated. Cumulative high active waste per ton of initial heavy metal is also calculated. Results show that the recycling of actinides and fission products in fast reactors through the atomics international reduction oxidation process results in a reduction of the potential hazard of radioactive waste.

Thermal Decomposition of Nitrated Tri-n-Butyl Phosphate in a Flow Reactor

Tri-n-butyl phosphate (TBP) is a universal nuclear extractant, commercially used in the PUREX process for the last 60 years. However, it is prone to nitration and thermal degradation, and as a consequence a red-oil event may be initiated under several operating conditions resulting in severe pressurization of vessel/cell if venting is inadequate. In this work, an attempt was made to understand the reaction pathway of thermal decomposition of nitrated TBP in a flow reactor at atmospheric pressure. Many reaction products were identified and quantified by instrumental methods like HPLC-RI and GC-TCD. The experimental data was analysed with a power law model and the apparent rate constants were estimated. The activation energy for thermal decomposition of nitrated TBP, assuming an Arrhenius type of temperature dependency, was estimated to be 47.39 ± 0.25 kJ·mol−1. The effect of both varying temperature and concentration of nitric acid on conversion of TBP into degradation products and products distribution was experimentally studied. Based on the experimental observations, a reaction mechanism framework for thermal decomposition of nitrated TBP is proposed.

Approach to on-line monitoring of PUREX process using chemometric processing of the optical spectral data

Abstract Optical spectroscopic measurements in the UV-Vis and IR ranges were performed in model solutions of aqueous and organic phases of the PUREX process for spent nuclear fuel (SNF) reprocessing. Chemometric processing of the spectral data with PLS (partial least squares) regression allowed simultaneous quantification of several key components (uranium, neptunium, plutonium, nitric acid) in these mixtures in an effective and elegant way. The content of all key components was quantitively determined with mean relative errors not exceeding 10%. It was shown that the employment of the whole spectra or their certain continuous regions for a PLS calibration enables to decrease the analytical errors compared to the use of a single wavelength in an ordinary least squares approach. The results of this research imply that the development of on-line techniques for SNF reprocessing monitoring is fully possible and can be based on optical spectroscopy methods combined with multivariate data processing techniques.

Comparative studies on the determination of di-n-butyl phosphate in degraded solvent of PUREX process by ion chromatography and gas chromatography methods

This paper describes comparative studies on the determination of di-n-butyl phosphate (DBP) by ion chromatography (IC) and gas chromatography (GC) techniques in spent solvent of PUREX process used for the reprocessing of spent nuclear fuels. The ion chromatography method involves the separation of DBP from 30% TBP–NPH (tri-n-butylphosphate diluted in normal paraffin hydrocarbon) containing heavy metal ion like uranium and nitric acid by extraction of DBP into alkaline medium. DBP was subsequently eluted by ion-exchange separation in ion chromatography column and followed by suppressed conductivity detection. DBP is quantified to a lower limit of about 1 ppm with 3% RSD. However, in order to determine DBP by gas chromatography technique DBP is first quantitatively converted into its volatile and stable derivatives by using diazomethane prior to analysis by GC. Results obtained with ion chromatographic technique are compared with those of obtained by standard gas chromatographic technique. It was observed that IC technique involves minimum steps and is much faster than GC analysis. The effect of mobile phase flow rate on the detector response, retention time, and column pressure are also investigated.