Separation, for Analytical Purpose, of Np Traces from different Solutions of Fuel Reprocessing

AbstractFour separation methods were developed for performance control of hydrometallurgical extraction processes as COEX™ or advanced PUREX. These methods used implemented the operations of radionuclides oxidation state adjustment and chromatographic separation using TEVA resin. Concerning FP raffinate, the method consisted in reducing Np traces to the valence IV by a mixture of ferrous sulfamate and ascorbic acid, to fix Np(IV) on “TEVA” resin and to eluate it by a nitrohydrofluoric acid solution. The Np recovery yield is 100%. The decontamination of Np is sufficiently high to allow its analysis by FXL (Zr/Np < 1). The study also showed that in presence of Zr and Tc, Pu behaved like Np. The mixture of ferrous sulfamate and ascorbic acid had surprisingly no action on Pu(IV). Concerning plutonium solution ([Pu] > 10g/L) and uranium solution ([U] > 100g/L), the same method used for Np recovery from FP raffinate led to an eluate containing 100% of the initial Np ([Np]: 10mg/L). The low concentration of U and Pu (< 100mg/L) allows the determination of Np by FXL. Concerning Pu(III)-U(IV) solution, the method, included 2 redox stages, the first one to oxidize all actinides to oxidation state VI et the second one to reduce Np and Pu respectively to IV and III oxidation state. Then Np(IV) was fixed on TEVA resin. The eluate contains 100% of the initial Np ([Np]: 10mg/L) and a low concentration of U and Pu ([U] < 20mg/L, [Pu] < 10mg/L). The next experiments will consist in consolidating these good results by working with real solutions of fuel reprocessing

Actinide L-line ED-XRF and HKED Spectra Processing

International audienceThe analysis laboratory in the CEA Atalante facility at Marcoule (France) performs numerous R&Dstudies carried out in glove-boxes or in hot cells. The samples are measured in liquid phase,aqueous or organic.The concentration of the main actinides of interest (U, Np, Pu, Am and Cm) are determined by XRFin a hot cell using a device built specifically for these actinides analyses via their L-line X-raybetween 13 and 15 keV.For sample with U and/or Pu in high concentrations, the hybrid K-edge densitometer is used.New software were developed for these devices

Separation of amercium from a concentrated raffinate by liquid-liquid extraction hot tests in the ATALANTE facility

National audienceAbstract – Recycling americium (Am) alone from the spent fuel is an important option studied forthe future nuclear cycle. Since 2008, a liquid-liquid extraction process called EXAm has beendeveloped by the CEA to allow the recovery of Am alone from a PUREX or COEXTM raffinate(already cleared from U, Np and Pu). A mixture of DMDOHEMA (N,N'-dimethyl-N,N'-dioctyl2-(2-(hexyloxy)ethyl)-malonamide) and HDEHP (di-2-ethylhexylphosphoric acid) in TPH is usedas the solvent and the Am/Cm selectivity is improved using TEDGA (tetraethyldiglycolamide) as aselective complexing agent to maintain Cm and heavier lanthanides in the acidic aqueous phase(HNO3 5M). Americium is then stripped selectively from light lanthanides at low acidity (pH2.5-3)with a polyaminocarboxylic acid.Since 2011, in order to increase the compactness of this process and future plan associated,additional developments have been studied to adapt the EXAm process to a concentrated raffinate(addition of a TEDGA stripping step, pH in the Mo stripping step more difficult to control,development of a model in low acidic conditions). Following up first cold tests in G1 facility in2011-2012, a test was carried out in April 2014 in the ATALANTE facility (C17 hot cells) on asurrogate feed solution with trace amounts of americium and curium. This C17 test aimed atconsolidating the process flowsheet and ensuring performances achievement in terms ofamericium recovery and its decontamination towards curium and light lanthanides.An americium recovery rate of 99.3% was obtained. Although the americium flow was notsufficiently decontaminated towards Cm and light lanthanides, several ways of improvement wereidentified. A simulation of this test was performed with the PAREX code and compared withexperimental data, either in transitory situations or at the end of the test. The modelling of mainelements was broadly validated.Based on this work, a flowsheet is proposed for a hot test on a genuine PUREX raffinatescheduled in 2015 in CBP hot cells in ATALANTE facility. The recovery of several grams ofamericium is expected to produce some AmO2 pellets for irradiation experimentations

Consolidation of the EXAm Process: Towards the Reprocessing of a Concentrated PUREX Raffinate

International audienc

Decontamination of high level activity solutions from cesium and strontium by chromatography

International audienceThe high rate of irradiation of certain samples resulting from various programs carried out in ATALANTE facility is essentially due to the presence of $^{137}$+$^{134}$Cs and $^{90}$Sr. The elimination of these radionuclides would make it possible to handle these samples in glove boxes without having to carry out important dilution beforehand. It would allow to analyze these samples by techniques usually implemented in glove boxes (ICP, radiometry, ...) and to reach lower detection limits.Preliminary tests were carried out with real samples by using extraction chromatography as separative technique.The principle of the cations/Cs separation consists in adding on a 2 mL column filled with resin AM-PAN (ammonium molybdophosphate embedded in polyacrylnitrile), the sample adjusted before at a nitric acidity of 2M, then a washing solution ([HNO$_3$] = 2M). During these two steps, Cs$^+$ is expected to be retained by the resin while the other cations cross the column.This separation, tested on an aliquote of a solution of irradiated fuel dissolution, showed that contrary to the cesium, elements measured by ICP-AES (U and Pu) and by gamma spectrometry ($^{154}$+$^{155}$Eu, $^{244}$Ce, $^{241}$Am and $^{106}$Ru) are not retained during the feeding and washing steps. The decontamination factor of cesium is superior to 500.The principle of the cations / Sr separation consists in adding on a 2 mL column filled with a Sr resin ((1.0M 4,4 ' (5 ')-di-t-butylcyclohexano 18-crown-6 (crown ether) in 1-octanol (weighing 40%) / inert chromatographic support (weighing 60%)) the sample, adjusted before at the acidity [HNO$_3$] = 3M / [H$_2$C$_2$O$_4$] = 0.01M, then a washing solution ([HNO$_3$] = 3M / [H$_2$C$_2$O$_4$] = 0.01M). During these two steps, the strontium is expected to be retained by the resin whereas the other cations cross the column.This separation, tested on an aliquote of a solution of irradiated dissolution of fuel, showed that elements measured by ICP-AES (U, Fe, Mo, Zr, Pd, Zr, Pd, Ru, Tc) and by spectrometry gamma ($^{134}$+$^{137}$Cs, $^{154}$+$^{155}$Eu, $^{241}$Am, $^{106}$Ru) are not retained during the feeding and washing stages. The passage on the column of an elution solution ([HNO$_3$] = 0.01M) led to the total recovery of the initial strontium. The results of the cations / Cs and cations / Sr separations are very encouraging. Additional experiments on other high level radioactive samples are in progress to consolidate both tested methods. They consist in using a wider range of samples, studying the influence of fluorides and determining the behavior of a largest number of elements. These studies will be presented on the poster

Demonstration of uranium - plutonium separation and purification from spent nuclear fuel with monoamide solvent

International audienceIn the framework of the development of Generation IV reactors, innovative solvent extraction processes are under development for the reprocessing of spent nuclear fuels. Monoamides demonstrated their potentiality in the recovery and recycling of fissile materials, plutonium and uranium, as an alternative to TBP. First, they exhibit a good stability towards radiolysis and hydrolysis. Secondly, distribution ratios of Pu(IV) and U(VI) with monoamides are such that their extraction and separation is possible, without using any reducing agent for the uranium - plutonium partitioning. These potentialities were demonstrated during pilot tests performed on a genuine High Liquid Waste (HLW) in the CBP hot cell (Atalante facility). The HLW was obtained from the dissolution of irradiated uranium oxide fuels with burnup between 25 to 65 GWd/t. The contactor set-up consisted of six batteries of mixer-settlers. The first three mixer-settlers banks were devoted to the uranium and plutonium extraction and to the fission products scrubbing. The two following batteries were dedicated to the uranium – plutonium partitioning. The last step consisted in the stripping of uranium. In addition to the six mixer-settlers banks, the solvent clean-up was carried out thanks to three centrifugal contactors, allowing its recycling. After running at least 70 hours, more than 99.99% and 99.97% of respectively the initial uranium and plutonium were recovered with high decontamination factors versus fission products (mainly 99Tc, 106Ru and 137Cs). The organic and aqueous concentration profiles of uranium in the different stages of the process were analysed and these experimental data were compared with the calculated values. The comparison between experimental and predicted concentration profiles exhibits a good agreement between the two sets of data

Lab-scale implementation of a next-generation uranium and plutonium separation and purification process from spent nuclear fuel using monoamide solvent

International audienc