Electrorefining of metallic spent nuclear fuel (SNF) has been demonstrated using irradiated fuel from Experimental Breeder Reactor-II at Idaho National Laboratory and is being considered for recycling SNF. Metal-chloride reduction potentials are key to predicting constituent metal behavior during SNF electrorefining. Rare earths are of particular interest, as major fission products in SNF are known to have very close equilibrium reduction potentials. Different sources of data on these reduction potentials indicate conflicting sequences. A comparative study was performed in which Gd, Nd, Ce, and La metal rods were equilibrated in molten eutectic LiCl-KCl with 1 wt% LaCl3 at 773 and 873 K. Equilibrium potentials of Gd, Nd, and Ce were measured versus La/La & sup3;(+), and concentrations of GdCl3, NdCl3, and CeCl3 were measured via inductively coupled plasma mass spectrometry. All reported equilibrium potentials and derived activity coefficients are relative to the La/La & sup3;(+) redox couple. The order of equilibrium potentials was Gd > Ce > Nd > La. Measured potential differences ranged from 2.76 to 32.12 mV. Relative activity coefficients for GdCl3, NdCl3, and CeCl3 with respect to that of LaCl3 were calculated and varied from 0.18 to 11.68. The order, by magnitude, of the activity coefficients is CeCl3 > LaCl3 > GdCl3 approximate to NdCl3. This proved to be an effective method for measuring relative potentials and activity coefficients.

This paper presents comparative volatility studies for the two molten salts researched as surrogate salts to FLiBe; FLiNaK and FLiNaZr. It was found that initial mass loss of FLiNaZr is dominated by water evaporation and hydrolysis to form HF. After this initial stage, the two salts exhibit comparable rates of vaporization at 700 °C, during long-term applications. Methods of experimentation included isothermal thermogravimetric analysis (TGA), transpiration in a tube furnace, and long-term isothermal heating in a muffle furnace, open to a glove box atmosphere. Partial pressures of constituent metal fluorides in both salts were calculated using data from the TGA/transpiration experiments.

A method combining thermogravimetric analysis (TGA) and horizontal transpiration with elemental analysis via inductively coupled plasma mass spectrometry or ion chromatography enabled calculation of partial pressures of individual salts in molten mixtures. TGA quantified total mass loss, while transpiration identified vapor-phase composition. Two chloride (NaCl-MgCl2, NaCl-MgCl2 + UCl3) salts and one mixed halide (LiCl-LiF + Li2O) salt were analyzed at 750 °C and 550 °C, respectively. NaCl and MgCl2 vapor pressures were 2.19–2.61 × 10-4 atm and 2.47–2.48 × 10-5 atm (dependent upon the identity of the invesitgated mixture); UCl3 was 1.42 × 10-7 atm. LiCl and LiF vapor pressures at 550 °C were 1.53 × 10-6 and 6.32 × 10-6 atm, respectively. Additionally, the TGA method was validated against values from the literature for unary LiCl and LiF.

Open circuit potential (OCP) measurements were made to determine uranium chloride (UClx) species activity in NaCl-MgCl2, which is a candidate salt for use in a molten salt reactor fuel. The operating temperature ranged from 500 to 650( degrees)C, and the measurement system used a U-Zr working electrode (WE) and Ag/AgCl reference electrode (RE). The activity was calculated to range from 2.52x10(-6) to 2.99x10(-8) (with uncertainties of +/- 1.0x10(-7) and +/- 1.1x10(-9), respectively), dependent upon the temperature and concentration of UCl3 (ranging from 0.25 to 11.55 wt.%). Fit of the OCP data to the Nernst equation resulted in values of 3.1 to 4.8 electrons transferred per U atom. This indicates the possibility for codeposition of Mg2+ and U3+ under certain conditions. Scanning electron microscopy, energy-dispersive X-ray spectroscopy, and X-ray photo-electron spectroscopy analyses of cathodic deposits confirm the presence of metallic magnesium.

In order to meet the safeguard requirements for Nuclear Material Control and Accountability (MC&A) in liquid fueled molten salt reactors, fuel salt transportation, and fuel cycle processes, it is important to develop effective methods for measuring and monitoring the mass of molten fuel salt or coolant salt. This paper seeks to validate the radioactive tracer dilution (RTD) method used for determining the mass of irradiated uranium-bearing molten salt. Small scale testing was conducted for this method by irradiating 6.065 g of MgCl-KCl-UCl3 salt mixture with 1.21 mg of 235U at The Ohio State University, resulting in a neutron fluence of 4.0 x 1016 n/cm2. Gamma spectra were acquired over the course of a 42-day period starting in March 2023 to assess fission products and activity levels of the 22Na tracer that was added before irradiation. No additional interference peaks were observed to overlap with the 1274.54 keV energy peak from 22Na, aside from the known 154Eu peaks, which can be corrected based on prior studies. By placing a thin piece of lead between the source and detector, deadtime was effectively reduced while still allowing observation of the higher-energy peaks, most notably around the 154Eu peak at 1004.76 keV used for interference correction. Overall, the uranium salt irradiation techniques applied in this experiment demonstrate the potential of the RTD method to effectively measure small irradiated samples for large scale molten fuel salt applications.

Electrochemical processing of spent nuclear fuel in molten chloride salts results in radioactive salt waste. Chlorine removal from the salt has been identified as an effective and efficient first step in the management of high-level waste. In this work, a simple salt was dechlorinated with a phosphoric acid phosphate precursor, resulting in a glassy dechlorinated product. The dechlorination efficacy was evaluated in air and argon environments. This work serves as an initial step to advance the Technological Readiness Level of H3PO4-based dechlorination step toward implementation of iron phosphate waste forms to immobilize electrochemical fuel reprocessing salt waste streams.

In this study, an experimental high-temperature fluoride salt reference electrode (RE) developed by HiFunda LLC was tested in molten FLiNaK at 550 degrees C. The high-temperature reference electrode (HTRE), based on the Ni/NiF2 redox couple, was tested over a period of 13 d for short-term stability, long-term stability, and electrochemical analysis capabilities. The HTRE was tested using open circuit potentiometry (OCP), cyclic voltammetry (CV), square wave voltammetry (SWV), and electrochemical impedance spectroscopy (EIS). The HTRE reference potential was measured against changes in salt composition by increasing FeF2 concentrations in the melt across six additions from 0 to 0.1 mol% FeF2. The long-term stability of the electrode was then tested over ten days at a constant composition of 0.1 mol% FeF2. OCP, SWV, and CV were used to calculate an average potential drift between 7-10 mV per day. Using CV and SWV, the number of electrons transferred for iron reduction and the diffusion coefficient of Fe2+ were measured.

Conducting research experiments on plutonium electrorefining is difficult due to the significant hazards and regulations associated with nuclear materials. Finding a surrogate for plutonium electrorefining studies would enable more fundamental research to be conducted. Potential surrogates were first identified by determining the physical properties required to conduct electrorefining at the same conditions commonly used in plutonium electrorefining, a molten metal and molten CaCl2 at 1123 K. Ce-CeCl3, In-InCl3, and Pb-PbCl2 were the only potential surrogates identified using these constraints. Sn-SnCl2 was also tested at these same conditions. More potential surrogates were identified by changing the matrix salt and operating temperature. This expanded the potential surrogate list to also include Zn-ZnCl2, Sn-SnCl2, and Bi-BiCl3. Zn-ZnCl2 was used with the LiCl-CaCl2 (65:35 mol%) eutectic at 773 K. Sn-SnCl2 and Bi-BiCl3 were used with the LiCl-KCl-CaCl2 (50.5:44.2:5.3 mol%) eutectic at 673-773 K. Ce electrorefining in molten CaCl2 resulted in a difficult to separate colloid mixture of Ce, Ca and Cl. Electrorefining rates for In in molten CaCl2 were too slow due to InCl3 volatilizing out of the molten salt. Only trace amounts of SnCl2 was retained in the CaCl2 at 1123 K resulting in impractical electrorefining rates. Zn metal product was successfully collected in the LiCl-CaCl2 eutectic molten salt, but the metal obtained did not coalesce into one piece. Sn and Bi were successfully electrorefined in the LiCl-KCl-CaCl2 eutectic molten salt and coalesced into product rings with high yields and coulombic efficiencies. While a surrogate could not be identified using the same conditions as plutonium electrorefining, two possible surrogates, Sn-SnCl2 and Bi-BiCl3, were found that could imitate the physical configuration (i.e., molten salt on top of molten metal) of plutonium electrorefining at a reduced temperature using the eutectic LiCl-KCl-CaCl2 salt at 673-773 K in place of CaCl2 at 1123 K.

Molten salt mixtures containing LiCl–KCl and NaCl–MgCl2 have been infused with UCl3 via reaction of U metal and FeCl2. The process starts with base salt (LiCl–KCl or NaCl–MgCl2) drying/purification using hydrochlorination via bubbling anhydrous HCl. An auto-titrator running in pH-stat mode was used to determine the point at which there is no net reaction with the salt. U metal is contained in a porous stainless steel basket as it is submerged in the molten salt. The byproduct Fe metal forms dendrites on the basket walls, allowing for simple separation from the molten salt. Based on analysis of salt samples using inductively coupled plasma mass spectroscopy, UCl3 yield of 90