## 1 Introduction  
Different waste management strategies for the handling of irradiated used nuclear fuel from nuclear power plants are followed worldwide by different countries [1]. While several countries (e.g., Germany and the United States) follow a direct disposal strategy (i.e., no further treatment of irradiated used nuclear fuel), other countries (e.g., France) follow a recycling strategy for U and Pu [2; 3; 4; 5]. In any strategy, waste will ultimately need to be disposed of in an underground repository. Such underground repositories are extremely rare with only a handful of repositories in operation worldwide (e.g., the Waste Isolation Pilot Plant (WIPP) in the United States for the disposal of defense waste), and a few sites under investigation or in the licensing process. Therefore, a most efficient use of repository space is preferable. French studies showed that the separation of minor actinides (MA: Np, Am, and Cm), especially Am, would reduce the total repository volume by up to a factor of seven [6; 7; 8]. The separation and recycling of U and Pu from used nuclear fuel is a mature technology, demonstrated on an industrial scale using the Plutonium Uranium Reduction Extraction (PUREX) process [9]. This process can also be adapted to allow the concurrent separation of Np [10; 11]. Separation of the trivalent actinides Am and Cm from PUREX raffinate solutions, on the other hand, is an ongoing research interest. Many hydrometallurgical process candidates have been tested, but no development has yet reached a higher technical readiness level above laboratory-scale testing with genuine fuel [5; 12; 13; 14; 15; 16]. Recent developments are aimed at single-cycle processes employing highly selective chemical systems for the difficult separation of the chemically similar trivalent actinides (An(III)) and lanthanides (Ln(III)), compatible with directly using the PUREX raffinate [16; 17; 18; 19; 20; 21; 22]. Furthermore, to be applied under the harsh conditions of used nuclear fuel recycling, innovative processes should be robust, use conventional and commercially available equipment and chemicals, and show good hydrolytic and radiolytic stability. For this purpose, the Actinide Lanthanide Separation Process (ALSEP) process was developed and proved to fulfill these requirements [18; 23; 24; 25; 26; 27; 28; 29]. The optimized ALSEP solvent is composed of 0.5 mol L\({}^{-1}\) 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester (HEH[EHP], Figure 1) and 0.05 mol L\({}^{-1}\)\(N\),\(N\),\(N^{\prime}\),\(N^{\prime}\)-tetra-(2-ethylhexyl)-diglycolamide (T2 EHDGA, Figure 1) in \(n\)-dodecane [27]. Co-extraction of Zr(IV) and Pd(II) is prevented using 0.05 mol L\({}^{-1}\)_trans_-1,2-diaminocyclohexane-\(N\),\(N\),\(N^{\prime}\),\(N^{\prime}\)-tetraacetic acid (CDTA, Figure 1) as a masking agent in the feed [30], whereas co-extracted Mo is scrubbed using 0.75 mol L\({}^{-1}\) acetohydroamic acid (AHA, Figure 1) and 0.175 mol L\({}^{-1}\) ammonium citrate at pH 3. The separation of An(III) from Ln(III) is achieved using 0.015 mol L\({}^{-1}\) diethylene-triamine-\(N\),\(N\),\(N^{\prime}\),\(N^{\prime}\),\(N^{\prime\prime}\)-pentaacetic acid (DTPA, Figure 1) and 0.2 mol L\({}^{-1}\) ammonium citrate at pH 2. Previous demonstration tests, however, showed some kinetics limitations of the chemical system, which could be overcome using tailored 3D-printed centrifugal contactors with an extended mixing zone, as well as difficulties with stripping of Ln(III) from the solvent [27]. The test described in the present paper sought to demonstrate the ALSEP concept using commercially available centrifugal contactors of a more standard design. The previously observed kinetic limitations were overcome by adjustment of the solvent composition (decreasing the HEH[EHP] concentration from 0.75 to 0.5 mol L\({}^{-1}\)) and operating the An(III) stripping stages at a slightly lower pH than in the previous investigations.