Oral Presentation International Solvent Extraction Conference 2025

Developments and implementation of the spent nuclear fuel recycling process PUMAS at the ATALANTE facility (121970)

Sylvain COSTENOBLE 1 , Solenne MICHAUD 1 , Amandine DUTERME 1 , Marc MONTUIR 1 , Fabien LENGRAND 1 , Vincent VANEL 1 , Pauline MOEYAERT 1 , Pierre SARRAT 1 , Steve JAN 1 , Christian SOREL 1 , Manuel MIGUIRDITCHIAN 1
  1. CEA, DES, ISEC, DMRC, Univ. Montpellier, Marcoule, FRANCE

An innovative solvent extraction process for the reprocessing of spent nuclear fuels is developed by CEA with the support of Orano and EDF. The PUMAS (Plutonium Uranium MonoAmide Separation) process is based on a monoamide-based solvent as a substitute to TBP (Tri-Butyl-Phosphate), currently used in the industrial PUREX process. Monoamides have been selected as they exhibit a good stability towards radiolysis and hydrolysis and a high selectivity for uranium and plutonium towards fission products [1-2]. In addition, this new extracting system allows a uranium/plutonium partitioning step without using any redox agent. The flowsheet can be divided into four different steps :

  • U(VI) & Pu(IV) coextraction at high nitric acidity. In this part, a fission products scrubbing section is devoted to the purification of U and Pu towards the FP. This is completed with a section at low acidity for U/U partitioning preparation.
  • U/Pu partitioning at low nitric acidity. This step is completed with an U scrubbing stage to adjust U/Pu ratio in the plutonium aqueous output.
  • Tc scrubbing by reduction with U(IV).
  • U stripping at very low nitric acidity.

This process has already been tested in the CBP high-level shielded process line in the ATALANTE facility with actual spent nuclear fuel solutions [3]. During this test, a mixture of two different monoamides was used and this experiment demonstrated the scientific feasibility of the process. Recently, a new single monoamide molecule [4] was proposed resulting from screening studies. Batch data were then acquired and experimental validations of each step of the process were performed in the ATALANTE facility. On the basis of experimental extraction isotherms and thermodynamic modelling of these data, process flowsheets were calculated with the PAREX+ simulation code. A series of different pilot tests were carried out in hotcells with laboratory-scale mixer-settlers using surrogate or genuine feed solutions in order to validate the hydrodynamical behaviour and the separation performances.

 

In this paper, the development and the implementation of the PUMAS process will be presented from extraction data to pilot tests with a focus on the different high-shielded facilities and the experimental conditions. Then, the main conclusions and experimental results will be given in comparison with the calculated values by the PAREX + code. Eventually, the main lines of the studies planned to scale the PUMAS process up to the industrial scale will described in conclusion.

  1. [1] P.N. Pathak, L.B. Kumbhare, V.K. Manchada, Solv. Extr. Ion. Exch., 19(1), 105-126 (2001)
  2. [2] N. Condamines, C. Musikas, Solv. Extr. Ion. Exch., 10, 69-100 (1992)
  3. [3] S. Costenoble, M.-J. Bollesteros, F. Antégnard, C. Sorel, V. Vanel, M. Montuir, M. Miguirditchian, X. Hérès, V. Boyer-Deslys, S. Grandjean « Lab-scale implementation of a next-generation uranium and plutonium separation and purification process from spent nuclear fuel using monoamide solvent », ISEC 2017, Miyazaki, Japan
  4. [4] M. Miguirditchian, H. Roussel, M. Montuir, C. Marie, P. Moeyaert, S. Costenoble, M.-J. Bollesteros, C. Sorel, T. Randriamanantena, F. Lamadie, S. Charton « Demonstration trials using a new monoamide-based solvent for spent nuclear fuel reprocessing », ISEC 2022, Göteborg, Sweden
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