D. Serrano-Purroy

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.

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).

DIAMEX Counter‐Current Extraction Process for Recovery of Trivalent Actinides from Simulated High Active Concentrate

Abstract The partitioning of trivalent actinides was demonstrated with a new version of the French DIAMEX (DIAMide EXtraction) process. A continuous counter‐current experiment using a 16‐stage centrifugal extractor battery was tested using 1 mol/L N,N′-dimethyl‐N,N′-dioctyl‐hexylethoxy‐malonamide (DMDOHEMA) in TPH as the extractant. A high active concentrate (HAC), obtained after concentration and denitration of a high active raffinate (HAR) with a concentration factor of 10, was used as a feed. Based on results from cold and hot batch extraction experiments and computer code calculations, a flowsheet was developed and a full test was carried out using a simulated HAC solution spiked with radionuclides (241Am, 244Cm, 152Eu, and 134Cs). In the DIAMEX process, five extraction stages were sufficient to obtain Am and Cm (feed/raffinate) greater than 5000 and for the coextracted lanthanides decontamination factors between 1100 and 4500. Co‐extraction of zirconium, molybdenum, and palladium was prevented by using oxalic acid and HEDTA. The back extraction comprising 4 stages was also efficient and the recoveries of actinides were greater than 99.8%, which can be further improved by a minor process flowsheet optimisation. The experimental steady‐state concentration profiles of important solutes were determined and compared with model calculations and good agreement was generally obtained.

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.

Towards an optimized flow-sheet for a SANEX demonstration process using centrifugal contactors

Abstract The design of an efficient process flow-sheet requires accurate extraction data for the experimental set-up used. Often this data is provided as equilibrium data. Due to the small hold-up volume compared to the flow rate in centrifugal contactors the time for extraction is often too short to reach equilibrium D-ratios. In this work single stage kinetics experiments have been carried out to investigate the D-ratio dependence of the flow rate and to compare this with equilibrium batch experiments for a SANEX system based on CyMe4-BTBP. The first centrifuge experiment was run with spiked solutions while in the second a genuine actinide/lanthanide fraction from a TODGA process was used. Three different flow rates were tested with each set-up. The results show that even with low flow rates, only around 9% of the equilibrium D-ratio (Am) was reached for the extraction in the spiked test and around 16% in the hot test (the difference is due to the size of the centrifuges). In the hot test the lanthanide scrubbing was inefficient whereas in the stripping both the actinides and the lanthanides showed good results. Based on these results improvements of the suggested flow-sheet is discussed.

Reducing uncertainties affecting the assessment of the long-term corrosion behavior of spent nuclear fuel.

Reducing the uncertainties associated with extrapolation to very long term of corrosion data obtainable from laboratory tests on a relatively young spent nuclear fuel is a formidable challenge. In a geologic repository, spent nuclear fuel may come in contact with water tens or hundreds of thousands of years after repository closure. The corrosion behavior will depend on the fuel properties and on the conditions characterizing the near field surrounding the spent fuel at the time of water contact. This paper summarizes the main conclusions drawn from multiyear experimental campaigns performed at JRC-ITU to study corrosion behavior and radionuclide release from spent light water reactor fuel. The radionuclide release from the central region of a fuel pellet is higher than that from the radial periphery, in spite of the higher burnup and the corresponding structural modifications occurring at the pellet rim during irradiation. Studies on the extent and time boundaries of the radiolytic enhancement of the spent fuel corrosion rate indicate that after tens or hundreds of thousands of years have elapsed, very small or no contribution to the enhanced corrosion rate has to be expected from α radiolysis. A beneficial effect inhibiting spent fuel corrosion due to the hydrogen overpressure generated in the near field by iron corrosion is confirmed. The results obtained so far point toward a benign picture describing spent fuel corrosion in a deep geologic repository. More work is ongoing to further reduce uncertainties and to obtain a full description of the expected corrosion behavior of spent fuel.

Leaching of 53 MW/d kg U spent nuclear fuel in a flow-through reactor

Abstract The dissolution behaviour of powdered commercial spent fuel (UO2 with burn-up of 53 MW/d kg U) has been studied in a carbonate-containing solution ([HCO3-] =0.001 mol dm-3) by using a flow-through reactor specially designed for the use in a hot cell. This method allows studying spent fuel dissolution while avoiding the parallel process of secondary solid phase formation. The dissolution behaviour of U, Np, Pu, Sr and Cs was studied. The main trend of the results obtained in this work is that only neptunium releases congruently with uranium (FIAPNp/FIAPU=1.21±0.01) because both strontium and caesium have higher FIAP values (FIAPSr/FIAPU=2.3±0.8; FIAPCs/FIAPU=5±1) and plutonium lower (FIAPPu/FIAPU=0.07±0.02). The FIAP value for uranium at the steady-state is 4(±2)×10-4.

RN Fractional Release of High Burn-Up Fuel: Effect of HBS and Estimation of Accessible Grain Boundary

The so-called Instant Release Fraction (IRF) is considered to govern the dose released from Spent Fuel repositories. Often, IRF calculations are based on estimations of fractions of inventory release based in fission gas release [1]. The IRF definition includes the inventory located within the Gap although a conservative approach also includes both the Grain Boundary (GB) and the pores of restructured HBS inventories. A correction factor to estimate the fraction of Grain Boundary accessible for leaching has been determined and applied to spent fuel static leaching experiments carried out in the ITU Hot Cell facilities [2]. Experimental work focuses especially on the different properties of both the external rim area (containing the High Burn-up Structure (HBS)) and the internal area, to which we will refer as Out and Core sample, respectively. Maximal release will correspond to an extrapolation to simulate that all grain boundaries or pores are open and in contact with solution. The correction factor has been determined from SEM studies taking into account the number of particles with HBS in Out sample, the porosity of HBS particles, and the amount of transgranular fractures during sample preparation.