Jean-Paul Glatz

Demonstration of a TODGA based Extraction Process for the Partitioning of Minor Actinides from a PUREX Raffinate

Abstract: Efficient recovery of minor actinides (MA) from genuine PUREX raffinate has been successfully demonstrated by the TODGA + TBP extractant mixture dissolved in an industrial aliphatic solvent TPH. The process was carried out in centrifugal contactors using an optimized flow‐sheet involving a total of 32 stages, divided into 4 stages for extraction, 12 stages for scrubbing and 16 stages for back‐extraction. Very high feed decontamination factors were obtained (Am, Cm ∼ 40 000) and the recovery of these elements was higher than 99.99%. Of the non‐lanthanide fission products only Y and a small part of Ru were co‐separated into the product fraction together with the lanthanides and the MA.

Demonstration of a SANEX Process in Centrifugal Contactors using the CyMe4‐BTBP Molecule on a Genuine Fuel Solution

Efficient recovery of minor actinides from a genuine spent fuel solution has been successfully demonstrated by the CyMe4‐BTBP/DMDOHEMA extractant mixture dissolved in octanol. The continuous countercurrent process, in which actinides(III) were separated from lanthanides(III), was carried out in laboratory centrifugal contactors using an optimized flow‐sheet involving a total of 16 stages. The process was divided into 9 stages for extraction from a 2 M nitric acid feed solution, 3 stages for lanthanide scrubbing, and 4 stages for actinide back‐extraction. Excellent feed decontamination factors for Am (7000) and Cm (1000) were obtained and the recoveries of these elements were higher than 99.9%. More than 99.9% of the lanthanides were directed to the raffinate except Gd for which 0.32% was recovered in the product.

Electrochemistry of Uranium in Molten LiCl-KCl Eutectic

The Institute for Transuranium Elements ~ITU! is building up an accurate database of actinide behavior in chloride melts in support of its nuclear fuel reprocessing development program. The electrochemical properties of uranium, dissolved in LiCl-KCl eutectic melt, were investigated by transient electrochemical techniques, such as cyclic voltammetry ~CV! and chronopotentiometry on an inert tungsten electrode. It was shown that U4+ is reduced to U0 by a two-step mechanism corresponding to U4+/U3+ and U3+/U0 transitions. In the 400-550°C s673-823 Kd range, the diffusion coefficients of U3+ and U4+ were similar and equal to: DsU3+d = 13.7 3 10−4 exph−24.2 3 103/RTsKdj and DsU4+d = 5.25 3 10−4 exph−19.8 3 103/RTsKdj cm2 s−1. The apparent standard potentials of U4+/U3+ and U3+/U0 redox systems were Eo*sU4+/U3+dsVd = −1.902 + 0.0006104T sKd vs. Cl2/Cl and Eo*sU3+/U0dsVd = −3.099 + 0.0007689T sKd vs. Cl2/Cl, respectively. Some thermochemical properties sDG*,gd of uranium solutions were also derived from the electrochemical measurements. The Gibbs free energies of dilute solution of UCl3 and UCl4 in the LiCl-KCl were determined to be: DG* = −897.09 + 0.226T sKd and −183.53 + 0.0589T sKd in kJ mol−1, respectively. In the 400-550°C s673-823 Kd range, the activity coefficients g of UCl3 and UCl4 range between 0.3 − 4.5 3 10−3 and 17.1 − 12.7 3 10−3, respectively.

Electroseparation of Actinides from Lanthanides on Solid Aluminum Electrode in LiCl-KCl Eutectic Melts

University of Grenoble, ENSEEG, 38402 Saint Martin d’He`res, FranceThis work presents a study on the electrochemical behavior of actinides (An 5 Am and Pu! and lanthanides (Ln 5 La and Nd!onto solid aluminum cathodes in a molten LiCl-KCl eutectic at 733 K. Cyclic voltammetry of these elements onto Al workingelectrode is carried out to estimate the reduction potentials of An and Ln and to predict the efficiency of an An/Ln separation byelectrolysis. Results show that the reduction of Am

Recovery of Minor Actinides from HLLW Using the DIAMEX Process.

Summary In this work a hot demonstration of the DIAMEX process using the new reference molecule DMDOHEMA (N,N′-DiMethyl-N,N′-DiOctylHexylEthoxyMalonAmide) and genuine high-level PUREX raffinate as feed is reported. The continuous counter-current experiment was carried out in a 16-stage centrifugal extractor battery, installed in a hot cell. In order to produce a representative High Level Liquid Waste (HLLW) a PUREX process was applied on dissolved fuel using the same equipment. It was demonstrated that 5 extraction stages were sufficient to achieve feed decontamination factors above 2200 and 320 for Am and Cm, respectively. Co-extraction of molybdenum and zirconium, as well as of palladium, were efficiently prevented using oxalic acid and HEDTA scrubbing, respectively. However, technetium, yttrium, and to some extent ruthenium, were co-extracted. The back-extraction proved to be very efficient, yielding in 4 stages more than 99.9% recovery of Am and Cm. The extraction profiles were modeled with computer code simulations and the results compared with data obtained from an experiment using the former reference molecule DMDBTDMA (DiMethylDiButylTetraDecylMalonAmide).

Recent achievements in the development of partitioning processes of minor actinides from nuclear wastes obtained in the frame of the Newpart European Programme (1996-1999)

Abstract Partitioning of long-lived minor actinides (americium and curium) from the nuclear wastes issuing the reprocessing of nuclear spent fuels, in order to transmute them into short-lived nuclides or to condition them into stable crystalline matrices, was the subject of intense research within the NEWPART research program of the European 4 th Frame Work Program, FWP (1996–1999). The target waste considered was the acidic raffinate (HAR) issuing the reprocessing of the used nuclear fuels by the PUREX process. A two step separation process based on liquid-liquid extraction was designed. The first step consists in the co-separation of the mixture of trivalent actinides and lanthanides from the HAR by extraction with a malonamide extractant (DIAMEX process), while the second step concerns the actinides(III)/lanthanides(III) group separation (SANEX process). Several DIAMEX and SANEX processes were developed and successfully tested with cold, spiked and genuine high active effluents. The research carried out also included basic and fundamental works in order to better understand the relationships between the structures of the extractants and their affinities for the target metal ions. The lecture highlighted both the basic and applied aspects of the research. This work will be pursued (PARTNEW program) within the 5 th FWP of the European Union during the period 2000–2003.

Promising Pyrochemical Actinide/Lanthanide Separation Processes Using Aluminum

Abstract Thermodynamic calculations have shown that aluminum is the most promising metallic solvent or support for the separation of actinides (An) from lanthanides (Ln). In molten fluoride salt, the technique of reductive extraction is under development in which the separation is based on different distributions of An and Ln between the salt and metallic Al phases. In this process molten aluminum alloy acts as both a reductant and a solvent into which the actinides are selectively extracted. It was demonstrated that a one-stage reductive extraction process, using a concentrated solution, allows a recovery of more than 99.3% of Pu and Am. In addition excellent separation factors between Pu and Ln well above 103 were obtained. In molten chloride media similar separations are developed by constant current electrorefining between a metallic alloy fuel (U60Pu20-Zr10Am2Nd3.5Y0.5Ce0.5Gd0.5) and an Al solid cathode. In a series of demonstration experiments, almost 25 g of metallic fuel was reprocessed and actinides collected as An-Al alloys on the cathode. Analysis of the An-Al deposits confirmed that an excellent An/Ln separation (An/Ln mass ratio = 2400) had been obtained. These results show that Al is a very promising material to be used in pyrochemical reprocessing of actinides.

Advanced Aqueous Reprocessing in P&T Strategies. Process Demonstrations on Genuine Fuels and Targets.

Summary In the present work the performance of several processes used for advanced reprocessing of commercial LWR fuels as well as transmutation targets is compared. As a first step uranium and plutonium were recovered by PUREX type reprocessing. The raffinate, containing fission products including lanthanides and the minor actinides (MA) was used as feed for the second step in which minor actinides and lanthanides were separated from the bulk of the fission products. The five different processes tested use CMPO, DIDPA, TRPO, Diamide and CYANEX 923 as extractants. In the third step MA are separated from lanthanides. Here three processes were tested, i.e. using CYANEX 301, the synergistic mixture of di-chloro substituted CYANEX 301 and TOPO, and BTP solvents. Column-, batch- and continuous counter-current extraction techniques were used for the tests. The different processes will be described and discussed in terms of performances and efficiencies for Am and Cm separation. Efficient separation of MA from different genuine fuel solutions could be demonstrated and thereby also the possibility of closing a future transmutation fuel cycle. The combination of Diamide and BTP seems to be the best, among extractants tested, to achieve an efficient MA recovery from spent fuel.

Investigation of the radiolytic stability of a CyMe4-BTBP based SANEX solvent

Abstract The radiolytic degradation of the 6,6′-bis(5,5,8,8tetramethyl-5,6,7,8-tetrahydro-benzo[1,2,4]triazin-3-yl)-[2,2′]bipyridine (CyMe4BTBP) based SANEX (selective actinide extraction) solvent has been investigated. As the solvent used in the extraction process is designed to separate trivalent actinides from lanthanides, the radiolytic degradation is mainly due to alpha decay of extracted minor actinide isotopes. A calculation of dose-rates was done by estimating the concentration of minor actinides in the solvent by fuel burn-up calculations and assumptions on dilutions in the subsequent reprocessing steps. The calculations showed that the main isotopes responsible for the dose-rate are 242Cm, 244Cm and 241Am. 242Cm is short-lived and has an impact only at short cooling times before reprocessing of the spent fuel. The dose-rates to a SANEX solvent in the reprocessing of standard spent LWR fuels are burn-up dependent and range from at least 0.03–0.2 kGy/h for UO2 fuels and from 0.4 to 0.8 kGy/h for MOX fuels. Fast reactor fuels yield dose-rates over 1 kGy/h. Based on these results, several radiolysis experiments were carried out in order to compare the effect of low LET external gamma radiation (0.2 kGy/h) and internal alpha radiation with different dose-rates (0.05, 0.2 and 1.0 kGy/h). Significant radiolytic degradation was shown in the gamma radiolysis and in the alpha radiolysis experiment at a dose-rate of 1 kGy/h. These experiments were continued up to an absorbed dose ∼1200 kGy and >300 kGy, respectively. Comparing the alpha radiolysis results for 0.2 kGy/h and 1.0 kGy/h, up to an absorbed dose of ∼120 kGy, no significant difference in the degradation for the different dose rates could be seen. The radiolytic degradation rate for gamma radiation was 40% higher than for alpha radiation.

Electrorefining of U-Pu-Zr-alloy fuel onto solid Aluminium cathodes in molten LiCl-KCl

An electrorefining process in molten chloride salts using solid aluminium cathodes is being developed at ITU to recover actinides (An) from the spent nuclear fuel. The maximum possible loading of aluminium electrodes with actinides was investigated during the electrorefining of U-Pu-Zr alloy in a LiCl-KCl eutectic at 450 °C. Two different electrolytic techniques were applied during the experiment and almost 6000 C has been passed, corresponding to 3.7 g of deposited actinides. A very high capacity of aluminium to retain actinides has been proven as the average Al: An mass ratio was 1:1.58 for galvanostatic and 1:2.25 for potentiostatic mode. The obtained deposits were characterized by XRD and SEM-EDX analysis and alloys composed of (U,Pu)Al3 were detected. The influence of zirconium co-oxidation during the process was also investigated and the presence of dissolved Zr ions in the melt yielded a significant deterioration of the quality of the deposit.