Recovery Of Actinides Research Articles

STATUS OF PYROPROCESSING TECHNOLOGY DEVELOPMENT IN KOREA

The Korea Atomic Energy Research Institute (KAERI) has been developing pyroprocessing technology for recycling useful resources from spent fuel since 1997. The process includes pretreatment, electroreduction, electrorefining, electrowinning, and a waste salt treatment system. This paper briefly addresses unit processes and related innovative technologies. As for the electroreduction step, a stainless steel mesh basket was applied for adaption of granules of uranium oxide. This basket was designed for ready handling and transfer of feed material. A graphite cathode was used for the continuous collection of uranium dendrite in the electrorefining system. This enhances the throughput of the electrorefiner. A particular mesh type stirrer was designed to inhibit uranium spill-over at the liquid Cd crucible. A residual actinide recovery system was also tested to recover TRU tracer. In order to reduce the waste volume, a crystallization method is employed for Cs and Sr removal. Experiments on the unit processes were tested successfully, and based on the results, engineering-scale equipment has been designed for the PRIDE (PyRoprocess Integrated inactive DEmonstration facility).

Extraction Behavior of Actinides and Metal Ions by the Promising Extractant, N,N,N′,N′-Tetraoctyl-3,6-dioxaoctanediamide (DOODA)

The extraction of actinides, fission products, some non-nuclear elements, and nitric acid by N,N,N′,N′-tetraoctyl-3,6-dioxaoctanediamide (DOODA-C8) in dodecane was extensively studied. Also studied was the extraction of HNO3 and Nd(III) by the tetradodecyl analog of DOODA-C8 in dodecane. Both extractants contain two ether oxygen atoms in the backbone chain carrying the two amide groups and can thus act as tetradentate ligands. The extractability of actinides decreases in the order Pu(IV) > U(VI), Am(III) > Np(V) in the extraction from nitric acid and Pu(IV) > Am(III) >> U(VI) in the extraction from perchloric acid. Ions of di-, tri-, tetra-, hexa-, and heptavalent metals strongly differ in the extractability by DOODA-C8 but, except for lanthanides(III), there is no visible correlation of their distribution ratios with ionic radii. Due to the efficient extraction of actinides, weak extraction of fission products, and sufficient extraction capacity, DOODA-C8 is a promising extractant for the recovery of minor actinides from high-level radioactive wastes.

Development of T2EHDGA Based Process for Actinide Partitioning. Part I: Batch Studies for Process Optimization

N,N,N′,N′-tetra-2-ethylhexyl diglycolamide (T2EHDGA) has been used for the extraction of actinides, lanthanides, fission products, and structural elements from Pressurized Heavy Water Reactor High Level Waste (PHWR-HLW). In view of third-phase formation under loading conditions, iso-decanol, tri-n-butyl phosphate (TBP), and N,N-di-n-hexyl octanamide (DHOA) were evaluated as the phase modifiers along with T2EHDGA. The loading of lanthanides (e.g. Nd(III), used as surrogate for trivalent minor actinides, Am(III) and Cm(III) and rare earth elements) in the organic phase sharply increased with the addition of iso-decanol as compared to other modifiers (TBP and DHOA). A minimum concentration of phase modifiers needed to avoid third-phase formation for 0.1 M T2EHDGA was 5% iso-decanol, 20% TBP or 20% DHOA. The most promising system, viz. 0.1 M T2EHDGA and 5% (v/v) iso-decanol was evaluated as extractant for actinide partitioning. The distribution behavior of various metal ions, viz. Am, Pu, U, Eu, Sr, Pd, Cs, Tc, Fe, and Mo has been studied from nitric acid as well as from synthetic PHWR-HLW using the optimized organic phase. Extraction of Am (tracer) from synthetic PHWR-HLW suggested a quantitative recovery of minor actinides in four contacts, while stripping (with 0.01 M HNO3) was quantitative in two contacts at O/A = 1. The testing of the optimized extraction system in mixer-settlers with simulated HLW is in progress.

Selective recovery of minor trivalent actinides from high level liquid waste by R-BTP/SiO2-P adsorbents

Concerning the selective recovery of minor trivalent actinides (MA(III) = Am(III) and Cm(III)) from high level liquid waste (HLLW) by extraction chromatography, adsorption and elution behaviours of MA(III) and fission products (FP) in a nitric acid media were studied using iHex-BTP/SiO2-P adsorbents, which is expected to show high adsorption affinity for MA(III) even in concentrated HNO3 solution, such as HLLW. In the batch experiments, Pd showed strong adsorption on iHex-BTP/SiO2-P adsorbents under any concentration of HNO3. The MA(III) and heavy Ln(III) (Sm(III), Eu(III) and Gd(III)) were also adsorbed at the condition of high HNO3 concentration, but they showed no adsorption under low HNO3concentration. The separation factor for MA(III)/heavy Ln(III) took the maximum value (over 100) at around 1mol/dm3 HNO3. It was difficult to elute MA(III) or heavy Ln(III) selectively by HNO3 from the iHex-BTP/SiO2-P adsorbents degradated by γ-ray irradiation. The chromatographic separation of real HLLW by an iHex-BTP/SiO2-P column showed that MA(III) could be recovered selectively by adjusting the acidity of the feed solution, i.e. HLLW, to 1mol/dm3 and using H2O as eluant. The adsorption of Pd(II) can be decreased by the addition of appropriate complexing reagents, e.g. DTPA, into HLLW without any effects on the MA(III) adsorption.

Sorption of Th, U, and Am on phosphorus-containing ion-exchange materials

Sorption of Th, uranyl, Am, Fe, and Al ions from nitric acid solutions on various phosphorus-containing ion-exchange materials (organic ion-exchange resins, SE-UPO solid extractant) was studied. The highest selectivity to quadruple-charged metal cations is exhibited by S-957 cation exchanger (Purolite) with phosphonic and sulfonic acid groups. The capacity of this resin for Th is essentially independent of the HNO3 concentration in the range 1–7 M and amounts to 80–95 mg ml−1. The uranyl, Am, Fe, and Al ions are sorbed on S-957 resin considerably worse than Th. The internal diffusion coefficient of Th4+ ions in S-957 resin from 3 M HNO3 was determined to be (5.3 ± 0.5) × 10−13 m2 s−1. The IR spectra of S-957 resin in the hydrogen and thorium forms were recorded. S-957 resin shows promise for selective recovery of tetravalent actinides from multicomponent nitric acid-salt solutions.

A study on the evaporation of cadmium for the recovery of actinides from a liquid cathode

The distillation behaviour of cadmium at a reduced pressure was investigated to develop an actinide recovery process from a liquid cadmium cathode in a laboratory scale cadmium distiller. The apparent evaporation rate of cadmium increased with an increasing temperature whereas the rate decreased with an increasing vacuum pressure. The evaporation rate of cadmium varied within 9.7–40 g/cm2/h in the temperature range of 500–650 °C and pressure range of 0.5–10 Torr (0.0667–1.33 kPa). The theoretical values calculated by the Hertz–Langmuir relation were much higher than experimentally measured values. The deviation was compensated by an evaporation coefficient (α) obtained empirically. About 0.02–0.20 wt% of residue was left in the crucible after distillation and found to be CdO. It could be concluded that the temperature range of 500–650 °C is favourable for the cadmium distillation process if residual eutectic salt does not exist in the cadmium alloy surface.

Recovery of Actinides and Lanthanides from High-Level Waste Using Hollow-Fiber Supported Liquid Membrane with TODGA as the Carrier

Transport behavior of trivalent actinides/lanthanides from nitric acid solutions into distilled water was investigated using polypropylene hollow-fiber supported liquid membrane (HFSLM) technique. A mixture of 0.1 M TODGA (N,N,N′,N′-tetraoctyl diglycolamide) + 0.5 M DHOA (N,N-dihexyl octanamide) dissolved in normal paraffinic hydrocarbon (NPH) was optimized as a suitable carrier. The permeability of the transported metal species was calculated with the help of various diffusional parameters. The permeation of metal ions increased with the feed acidity; however, their transport rates were significantly affected at higher acidity (>3 M HNO3) owing to carrier-aided co-transport of nitric acid. Mass transfer coefficients were calculated on the basis of the described model, whose success was verified by excellent matching of the membrane diffusion coefficient values calculated by the given model and Wilke−Chang equation. Quantitative transport of trivalent actinides and lanthanides was achieved in 30 min from ...

Sorption preconcentration of radionuclides on Taunit carbon nanostructural material

The possibility of using Taunit carbon nanostructural material for sorption recovery of radionuclides from aqueous solutions was examined. Taunit efficiently recovers Am, Pu, Eu, U, and Tc from weakly acidic and neutral solutions. Taunit is a suitable support for preparing solid extractants based on a phosphonium ionic liquid and diphenyl(dibutylcarbamoylmethyl)phosphine oxide for recovery of actinides from nitric acid solutions.

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.

Actinide Recovery Experiments with Bench-Scale Liquid Cadmium Cathode in Real Fission Product-Laden Molten Salt

This article summarizes the observations and analytical results from a series of bench-scale liquid cadmium cathode experiments that recovered transuranic elements together with uranium from a molten electrolyte laden with real fission products. Variable parameters such as the ratio of Pu3+/U3+ in the electrolyte, liquid cadmium cathode voltage, and feed materials were tested in the liquid cadmium cathode experiments. Actinide recovery efficiency and Pu/U ratio in the liquid cadmium cathode product under variable conditions are reported in this paper. Separation factors for actinides and rare earth elements in the molten LiCl-KCl/cadmium system are also presented.

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

: 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.

Actinide Recovery Experiments with Bench-Scale Liquid Cadmium Cathode in Real Fission Product-Laden Molten Salt

Actinide Recovery Experiments with Bench-Scale Liquid Cadmium Cathode in Real Fission Product-Laden Molten Salt

A study on the recovery of actinide elements from molten LiCl–KCl eutectic salt by an electrochemical separation

Pyroprocessing is a prominent way for the recovery of the long-lived elements from the spent nuclear fuel. Electrorefining is a key technology for pyroprocessing and generally composed of two recovery steps—deposition of uranium onto a solid cathode and the recovery of TRU (TRansUranic) elements. In this study, it was investigated on electrochemical separation of actinides to develop an actinide recovery system in a molten LiCl–KCl eutectic salt. In the electrorefining experiment, uranium was successfully separated from cerium. The effects of the anode material and the surface area were investigated during the electrolysis experiments for a more efficient electrorefining system. Anode potential decreased with an increasing anode surface area, however, an anode effect was observed in case of a complicated anode structure for high surface area. Glassy carbon was found to be the best anode material among the molybdenum, graphite, glassy carbon, and oxide materials. It was found that the solid cathode with a perforated ceramic container could be one of the candidates for a salt clean-up process to remove residual actinide elements in the salt after the recovery step.

Electrochemical Reduction of MOX Pellets in Molten Lithium Chloride Based on a Practical Operating Condition

Previous studies for electrochemical reduction using uranium oxide have shown that reduction was completed within several tens of hours when particles or powders of oxide were used for the cathode material. In the case of mixed oxide (MOX) fuel prepared for fast reactors, there are two significant differences with respect to uranium oxide fuel for light water reactors. The MOX fuel contains ~30% Pu and a small amount of Am. The density of the uranium oxide pellet and MOX pellet is ~98% and ~85% with respect to theoretical values, respectively. These differences decrease the electroconductivity of oxide and the reaction rate. Also, the behavior of transuranic elements has not been certified. In the present study, electrochemical reduction of MOX pellets was performed by setting the pellets directly on the cathode in a molten lithium chloride bath. Reduction was completed after ~15 h, even when using MOX pellets. This value compares closely to the previous values for uranium oxide particles or powders. Current efficiency was varied at ~60%, which is slightly higher than in the previous study. The lower density of MOX allows better diffusion of the molten salt into the pellet and contributes to efficient electrolysis. Concerning actinide behavior during electrolysis, the uranium and plutonium concentrations in the molten salt bath were lower than their detection limits. Although a small amount of americium was dissolved in the molten salt bath and gradually accumulated, the amount was <1% with respect to the initial amount. The oxygen concentration in the molten salt decreased gradually during electrolysis. These variations in the salt hardly affected the current efficiency and the actinide recovery ratio. These observations indicate that the electrochemical reduction of MOX pellets is applicable to industrial processes.

Solid extractants prepared with ionic liquids and their use for recovery of actinides from nitric acid solutions

Solid extractants were prepared by noncovalent immobilization of phosphonium ionic liquids (ILs) on various polymeric materials, including fibrous materials. These sorption materials can be used for recovering Pu from 0.5–5 M HNO3. Immobilization of phosphorus-containing ligands on solid supports using phosphonium ILs yields a material recovering actinides from nitric acid solutions.

Modification of a novel macroporous silica-based crown ether impregnated polymeric composite with 1-dodecanol and its adsorption for some fission and non-fission products contained in high level liquid waste

A novel macroporous silica-based chelating polymeric composite, DtDo/SiO 2–P, was synthesized by molecular modification of 4,4′,(5′)-di-( tert-butylcyclohexano)-18-crown-6 (DtBuCH18C6) with a long carbon chain organic compound 1-dodecanol. It was performed through impregnation and immobilization of DtBuCH18C6 and 1-dodecanol molecules into the pores of the SiO 2–P particles. The adsorption of a few fission and non-fission product elements Sr(II), Ba(II), Cs(I), Ru(III), Mo(VI), Na(I), K(I), Pd(II), La(III), and Y(III) onto DtDo/SiO 2–P was investigated at 323 K. The effects of contact time and the HNO 3 concentration in a range of 0.1–4.0 M were investigated. It was found that at the optimum concentration of 2.0 M HNO 3, DtDo/SiO 2–P exhibited strong adsorption ability and excellent selectivity for Sr(II) over all of the tested elements, which showed very weak or almost no adsorption except Ba(II). The bleeding of total organic carbon (TOC) from DtDo/SiO 2–P was evaluated. The quantity of TOC in aqueous phase increased with an increase in HNO 3 concentration in terms of a linear equation [TOC] = 35.82[HNO 3] + 115.5 with a correlation coefficient of 0.9751. The TOC content leaked from DtDo/SiO 2–P modified by 1-dodecanol, 119.0–269.3 ppm, in the range of 1.0–4.0 M HNO 3 was significantly lower than that of 424.8–634.6 ppm in the case without modification. It resulted from the intermolecular interaction force of DtBuCH18C6 and 1-dodecanol through hydrogen bonding. The reduction of DtBuCH18C6 leakage by molecular modification was achieved. It is of great benefit to application of DtDo/SiO 2–P in partitioning of Sr(II), one of the main heat generators, from high level liquid waste (HLLW) in reprocessing process of nuclear spent fuel in MAREC (Minor Actinides Recovery from HLLW by Extraction Chromatography) process developed.

Impregnation synthesis of a novel macroporous silica-based crown ether polymeric material modified by 1-dodecanol and its adsorption for strontium and some coexistent metals

4,4′,(5′)-Di-( tert-butylcyclohexano)-18-crown-6(DtBuCH18C6) is a chelating agent having high selectivity mostly for Sr(II). To significantly reduce its leakage by molecular modification, a macroporous silica-based DtBuCH18C6 polymeric composite (DtDo/SiO 2–P) was synthesized. It was performed by impregnating and immobilizing DtBuCH18C6 and 1-dodecanol molecules into the pores of the SiO 2–P particles utilizing an advanced vacuum sucking technique. The adsorption of a few fission and non-fission products Sr(II), Ba(II), Cs(I), Ru(III), Mo(VI), Na(I), K(I), Pd(II), La(III), and Y(III) onto DtDo/SiO 2–P was investigated. It was done by examining the effects of contact time and the HNO 3 concentration in a range of 0.1–5.0 M at 298 K. At the optimum concentration of 2.0 M HNO 3, DtDo/SiO 2–P exhibited strong adsorption ability and high selectivity for Sr(II) great over all of the tested elements, which showed very weak or almost no adsorption except Ba(II). Meanwhile, It was found that the quantity of total organic carbon (TOC) leaked from DtDo/SiO 2–P in 2.0 M HNO 3, 187.5 ppm, was lower than 658.4 ppm that leaked from DtBuCH18C6/SiO 2–P, which was not modified. This was ascribed to the effective association of DtBuCH18C6 and 1-dodecanol through intermolecular interaction. The reduction of DtBuCH18C6 leakage by molecular modification with 1-dodecanol was achieved. It was of great benefit to application of DtDo/SiO 2–P in chromatographic partitioning of Sr(II), one of the main heat generators, from high level liquid waste (HLLW) in reprocessing of nuclear spent fuel in the MAREC (Minor Actinides Recovery from HLLW by Extraction Chromatography) process developed recently.

On a fast reactor cycle scheme that incorporates a thoria-based minor actinide-containing cermet fuel

A fast reactor cycle scheme that incorporates a thoria-based minor actinide-containing cermet fuel is given. The present cermet fuel consists of an oxide solid solution of Th and minor actinides and Mo-inert matrix. It has been proposed as a high-performance device that can enhance minor actinide incineration in a fast reactor cycle. It is used in an independent small sub-cycle, whereby dedicated cycle technologies are adopted. Two-step reprocessing process was proposed for the present cermet fuel; it consists of a pre-removal of Mo-inert matrix and an actinide recovery. A preliminary test for the pre-removal of Mo-inert matrix was carried out using a surrogate cermet fuel. Burnup characteristics of a fast reactor core loaded with the cermet fuel were investigated by using neutronic calculation codes. It was revealed that a heterogeneous composition of Mo-inert matrix in inner and outer cores may lead to an effective transmutation of minor actinides and a flattened power density. It was concluded that the present cermet fuel was potentially promising as a high-performance incineration device of minor actinides for fast reactors.

Engineering-Scale Liquid Cadmium Cathode Experiments

Recovery of transuranic actinides (TRU) using electrorefining is a process being investigated as part of the Department of Energy (DOE) Advanced Fuel Cycle Initiative (AFCI). TRU recovery via electrorefining onto a solid cathode is very difficult as the thermodynamic properties of transuranics are not favourable for them to remain in the metal phase while significant quantities of uranium trichloride exist in the electrolyte. Theoretically, the concentration of transuranics in the electrolyte must be approximately 106 greater than the uranium concentration in the electrolyte to produce a transuranic deposit on a solid cathode. Using liquid cadmium as a cathode contained within a LiCl-KCl eutectic salt, the co-deposition of uranium and transuranics is feasible because the activity of the transuranics in liquid cadmium is very small. Depositing transuranics and uranium in a liquid cadmium cathode (LCC) theoretically requires the concentration of transuranics to be two to three times the uranium concentration in the electrolyte. Three LCC experiments were performed in an Engineering scale elecdtrorefiner, which is located in the argon hot cell of the Fuel Conditioning Facility at the Materials and Fuels Complex on the Idaho National Laboratory. Figure 1 contains photographs of the LCC assembly in the hot cell prior to the experiment and a cadmium ingot produced after the first LCC test. Figure 1. Liquid Cadmium Cathode (left) and Cadmium Ingot (right) The primary goal of the engineering-scale liquid cadmium cathode experiments was to electrochemically collect kilogram quantities of uranium and plutonium via a LCC. The secondary goal was to examine fission product contaminations in the materials collected by the LCC. Each LCC experiment used chopped spent nuclear fuel from the blanket region of the Experimental Breeder Reactor II loaded into steel baskets as the anode with the LCC containing 26 kg of cadmium metal. In each experiment, between one and two kilograms of heavy metal was collected in the LCC after passing an integrated current over 500 amp hours. Analysis of samples from the liquid cadmium cathode ingots showed detectable amounts of transuranics and rare-earth elements. Acknowledgements K. B. Davies and D. M. Pace for the mechanical and electrical engineering needed to prepare the equipment for the engineering-scale liquid cadmium cathode experiments.

Contribution of the Khlopin Radium Institute to the development of nuclear explosion technology for preparing actinides

The main stages in the development of a process for preparing actinides by underground nuclear explosions in rock salt at the Galit field are considered. The data of experiments performed by researchers of the Khlopin Radium Institute in the course of the process development are analyzed. The results of detailed study of the cavities after underground nuclear explosions and properties of the resulting radioactive products reprocessed for recovery of actinides are presented.