Separation of Actinides from Rare Earth Elements by Electrorefining in LiC1-KC1 Eutectic Salt

Song

et al. 2014

J Radioanal Nucl Chem

Separation of rare-earth (RE) elements from minor actinides is difficult in pyroprocessing, especially for light REs (e.g., Nd, Ce, Pr, and La). Herein, U-10 wt%Zr fuel slugs containing 0, 3, and 7 wt%RE (RE: a rare-earth alloy comprising 53 % Nd, 25 % Ce, 16 % Pr, and 6 % La by weight) were prepared by gravity casting. Their casting soundness was characterized by gamma-ray radiography and immersion density measurement, while their microstructures and compositions were analyzed using scanning electron microscopy and energy-dispersive X-ray spectroscopy. Their actual compositions and impurities were confirmed using inductively coupled plasma atomic emission spectrometry. The compatibility of casting components and recovery of radioactive materials were evaluated based on the RE content.

Section: Introductionmentioning

confidence: 94%

Fabrication and evaluation of rare-earth-bearing fuel slugs for sodium-cooled fast reactors

Song

et al. 2014

J Radioanal Nucl Chem

“…The uranium is in the form of UO 2 ; hence, the reduction of UO 2 to U metal is being investigated here. The reduced species then undergo electrorefining, [5][6][7][8][9] where U product and other transuranic species are separated and recycled into new fuel.Studies of spent fuel oxides in molten salts have identified two possible reaction pathways, differing reduction potentials and varying distribution of residual oxide in the product. In a study by Hermann et al, 10 crushed spent fuel was loaded into a stainless steel basket and z E-mail: d.brett@ucl.ac.uk submerged in molten LiCl-1 wt% Li 2 O at 650 • C. A platinum anode and a Ni/NiO reference electrode were used.…”

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confidence: 99%

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confidence: 99%

Electrochemical Reduction of UO₂ to U in LiCl-KCl Molten Salt Eutectic Using the Fluidized Cathode Process

et al. 2016

The electrochemical reduction of UO 2 to U metal has been investigated in both Fluidized Cathode (FC) and Metallic Cavity Electrode (MCE) cell arrangements. Differences in the local concentration of O 2− where the reduction takes place influnces the reduction potential. The fleeting contact of UO 2 particle contact with the current collector in case of the FC results in much less O 2− buildup compared to MCE. Consequently, UO 2 reduction occurs over a range of potentials in the FC and may involve separate two 2-electron steps compared to one apparent 4-electron step in the MCE. It is proposed that there are three discrete periods during the FC reduction process. The first is an induction period during which reduced uranium particles gradually adhere to the tungsten current collector. The second is reduction associated with a rapid growth in electrode area and consequent increase in current. The third is a slower reduction of the remaining oxide in the melt. Complete reduction of metallic U is achieved at −2.2 V (vs. Ag/Ag + ) with an estimated faradaic current efficiency of >92%. There are significant drivers for pursuing the next generation of nuclear reactors. Generation IV reactors offer both advancements in reactor design and also feature fully integrated fuel reprocessing capabilities. There are six types of Generation IV nuclear reactors and power plants under development. These are: the very-high-temperature reactor (VHTR), the sodium-cooled fast reactor (SFR), the supercriticalwater-cooled reactor (SCWR), the gas-cooled fast reactor (GFR), the lead-cooled fast reactor (LFR), and the molten salt reactor (MSR). The key features of each are summarized in Table I. The fuel type used in the majority of these reactors is metal. Given that most current reactors employ metal oxide (MOX or UOX) fuel and the majority of legacy waste is also metal oxide, the conversion of metal oxides to metals is an important component in a future nuclear power flow sheet.High temperature molten salt reprocessing technology (pyroprocessing) 2 for spent nuclear fuel offers a range of advantages when compared to aqueous reprocessing techniques. This is due to: an improved proliferation resistance (local processing reduces the likelihood of material diversion, no separation of plutonium); using facilities with a smaller footprint; a shorter cooling period for irradiated fuel, and also improved criticality safety margins. Different process systems and salts have been studied in pyroprocessing, most of which are summarized by the Nuclear Energy Agency (NEA). Conceptual flow sheets for the pyrochemical treatment of used nuclear fuel are described by Argonne National Lab (ANL). 4 The spent nuclear fuel, in the form of oxide pellets, is decladded and chopped, then passed on for an electrolytic reduction step. The uranium is in the form of UO 2 ; hence, the reduction of UO 2 to U metal is being investigated here. The reduced species then undergo electrorefining, [5][6][7][8][9] where U product and other transuranic species are separated an...

“…LiCl-KCl molten salt serves as an electrolyte for the high temperature pyroprocessing 1,2 and other industrial chemical processes. 3 As a reprocessing technology, pyroprocessing for used nuclear fuel treatment has been well developed.…”

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confidence: 99%

“…A n+ (solvent) + ne − ← → A (electrode) [1] where n is the charge number of the actinide ion, and e represents electron. According to the Nernst's law, the electrode potential at equilibrium can be expressed by: [2] where E 0 A n+ /A is the standard potential of the A/A n+ and F is the Faraday constant; a A n+ is the activity and can be expressed by: a A n+ = γ A n+ x A n+ [3] where x A n+ and γ A n+ are the mole fraction and activity coefficient of A n+ in molten salt, respectively.…”

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confidence: 99%

Direct Calculation of Concentration-Dependent Activity Coefficient of UCl₃in Molten LiCl-KCl

Zhou

Zhang

2015

J. Electrochem. Soc.

It is believed that the thermodynamic properties, such as activity coefficient of UCl3 in molten LiCl-KCl, are concentration-dependent. However, most of the available experimental data now are limited to the extreme dilution cases. In the present study, thermodynamic integration (TI) method was applied to calculate the Gibbs energy of supercooled UCl3 and UCl3 in LiCl-KCl molten salt with different concentrations of UCl3 and then calculate the activity coefficient directly for different concentrations at a given temperature of 723 K. The calculated activity coefficients at low concentration cases were compared with available experimental measurements, and the comparison indicates the calculation results agree well with the literature. Most importantly, the present calculation shows that the activity coefficient of UCl3 in the molten salt is indeed a function of its concentration and increases with the mole fraction.