Uranium - plutonium separation trials from spent nuclear fuel in CBP shielded cell

International audienceIn the framework of the development of Generation IV reactors, innovative solvent extraction (SX) 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.To validate these potentialities, the CBP shielded cell is a unique facility to conduct all operations from the reception of spent fuel to the demonstrative pilot tests on genuine High Liquid Wastes (HLW). CBP hot cell docking equipments allow the reception of both LWR and FNR spent fuels transported either by RD15IIB, TN106 or more recently by IR100. HLW is obtained after batch dissolution of spent fuels operated in 8 to 16 litres dissolver. Solvent extraction processes can then be validated during trials conducted either in mixer-settlers banks or in pulsed columns. Mixer-settlers batteries are operated to demonstrate scientific feasibility of the SX process while pulsed columns are used to assess the technical feasibility. CBP shielded cell is indeed the world unique RandD hot cell allowing the operation of four meter high pulsed columns with HLW. In addition to these devices, the monitoring of experimental concentration is conducted by coupling in-line measurements (pH, temperature, spectrophotometry..) and deported analysis performed in the analytical hot cell CBA which is connected via a pneumatic transport line

Implementation of americium recycling demonstration from spent nuclear fuel in the highlevel shielded process line CBPin the ATALANTE facility

International audienceAs part of the radioactive waste management act of 2006, the minor actinides recovery from spent nuclear fuel is investigated for heterogeneous recycling in future fast neutron reactors. Within minor actinides, americium is, after plutonium, the main contributor to residual heat of long term radioactive wastes and Am recycling is a way to decrease their radiotoxicity. That is why, for nuclear fuel manufacturing and geological repository, it would be better to separate americium from curium. The experimental demonstration of americium recycling from spent nuclear fuel to (U,Am)O$_2$ pellets production is currently performed in the ATALANTE research facility of CEA Marcoule. The following steps were carried out in the high-level shielded process line (CBP)1-Reception and dissolution of spent nuclear fuel in order to obtain the feed solution, 2-Clarification of this feed solution by cross-flow filtration in a way to remove the insoluble dissolution residues,3-Implementation of a liquid-liquid extraction process to separate the main actinides uranium and plutonium, from the fission products and the minor actinides (americium and curium),4-Purification and concentration of the raffinate solution containing all the fission products and the minor actinides thanks to a steam distillation step,5-Americium selective extraction thanks to the EXAm process adapted to the concentrated raffinate in order to separate this minor actinide from curium and other fission products.The description of the successive steps will be detailed on the basis of the experimental processes and analytical results measured at each step

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