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The Ca–Br cycle is a promising method for efficiently producing hydrogen from water; however, it suffers from limitations inherent to gas–solid reactions. The cycle depends on the repeatable transformation of calcium bromide to calcium oxide and back to calcium bromide. The use of pure solids for these reactions would lead to rapid particle degradation and slow reaction kinetics. To circumvent these problems, a new reactor concept based on molten calcium bromide with dissolved calcium oxide has been proposed and developed. Preliminary experimental results indicate that the solubility of calcium oxide in calcium bromide at 800°C is at least 1.2wt%, a level that is expected to be high enough to make the proposed process work as designed. Early attempts to hydrolyze molten calcium bromide indicated that solid calcium oxide formation may inhibit reactivity, and thus injection of the moisture into the salt must be optimized. [Copyright &y& Elsevier]

Abstract Electrochemical methods have been developed to measure metal ion concentrations in spent nuclear fuel electrorefiner by correlating concentration with current response. Analytes UCl 3 and GdCl 3 were studied in LiCl-KCl eutectic solvent at 773 K. A mixture matrix, which consisted of varying concentrations of UCl 3 and GdCl 3 from 1 to 10 wt% and from 1 to 3 wt%, respectively, was tested. Optimized normal pulse voltammetry (NPV) measurements were made on this mixture matrix and used for the concentration correlations. The accuracy of the prediction was improved by including solution resistance compensation and minimizing the working electrode surface area. The presence of GdCl 3 showed no effect on the U3+ NPV reduction current, but high U concentrations affected the Gd3+ NPV reduction current. This is speculated to be due to increased migrational current in the molten salt mixture. The average relative measurement errors obtained were 1.6% and 2.7% for UCl 3 and GdCl 3 respectively. [ABSTRACT FROM AUTHOR]

The drawdown of actinides is an important unit operation to enable the recycling of electrorefiner salt and minimization of waste. A new method for the drawdown of actinide chlorides from LiCl-KCl molten salt has been demonstrated here. Using the galvanic interaction between the Gd/Gd(III) and U/U(III) redox reactions, it is shown that UCl 3 concentration in eutectic LiCl-KCl can be reduced from 8.06 wt.% (1.39 mol %) to 0.72 wt.% (0.12 mol %) in about an hour via plating U metal onto a steel basket. This is a simple process for returning actinides to the electrorefiner and minimizing their loss to the salt waste stream. [ABSTRACT FROM AUTHOR]

Options for reducing purified metal chlorides to metallic form were evaluated in support of the development of a metal purification process based on the conversion of impure metal via hydriding followed by chlorination. Metallothermic reduction of CeCl 3 was tested using Li, Ca, and Na as reductant metals with and without a molten salt solvent in MgO and Y 2 O 3 -coated MgO crucibles. The effects of added reductant metal form, crucible material and dimensions, and mixture hold times above Ce melting temperature were studied. Direct contact of Li and CeCl 3 was ineffective due to their immiscibility. A low-density alloy was formed between Li and Ce using Li in a LiCl solvent. Na was ineffective at reducing CeCl 3 to any measurable degree. The highest percent recovery (71.4 %) and consolidation of Ce was achieved when using Ca in a CaCl 2 -NaCl solvent with 99.8 % removal of CeCl 3 from the salt. Side reaction with MgO crucibles was found to be at least partially responsible for reduced Ce recovery. [ABSTRACT FROM AUTHOR]

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–UCl 3 salt mixture with 1.21 mg of 235U at The Ohio State University, resulting in a neutron fluence of 4.0 × 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. [ABSTRACT FROM AUTHOR]

Transparent system experimental studies have been performed to provide measurement and analysis of oxygen bubble distributions at different sparging rates in LiCl-KCl molten salt at 500°C using a high-speed digital camera and an oxygen sensor. The results reveal that bubble sizes and rise velocities increased with an increase in oxygen sparging rate. The bubbles observed were ellipsoidal in shape, and an equivalent diameter based on the ellipsoid volume was calculated, ranging from 0.00263 m to 0.00407 m. Results also show that the bubble equivalent diameters are normally distributed. A Fanning friction factor correlation was used to predict a bubble's rise velocity. The oxygen mass transfer coefficients were calculated using the oxygenation model; these values were on the order of 10-4 m/s and followed a decreasing trend corresponding to an increasing bubble size and sparging rate. The diffusivities were calculated based on two different approaches--one based on physics of the bubbles and the other on systematic properties. The diffusivity values calculated from bubble physics are 1.65 x 10-9 m²/s to 8.40 x 10-4 m²/s, which are within the range suggested by literature for gases in liquids of a similar viscosity. [ABSTRACT FROM AUTHOR]

Abstract: Experimentation, data analysis, and modeling of ion exchange kinetics between fission products (cesium and strontium) and zeolite-A beads in molten LiCl–KCl have been performed to support optimization of an electrochemical process to treat used nuclear fuel. Models based on pseudo-first- and pseudo-second-order sorption as well as diffusion have been adapted and compared to experimental data to assess their validity in describing the system. Individual experiments were performed with different concentrations of CsCl or SrCl2 in the LiCl–KCl salt. Zeolite beads were removed from the molten salt at prescribed intervals of time and prepared for analysis by inductively coupled plasma-mass spectrometry. Results indicate maximum cesium (Cs) and strontium (Sr) loading occurred at approximately 31min and 104min of contact, respectively. The rate of loading and maximum loading were found to increase with increasing initial concentration of Cs or Sr. Data analysis included determination of rate constants and diffusion coefficients of the proposed models for each experimental condition. Results reveal that the diffusion model provides the best fit to the experimental data with average diffusion coefficients of 2.0×10−10 m2 s−1 for Cs and 6.3×10−11 m2 s−1 for Sr. This suggests that chemical diffusion is the dominant mechanism of mass transfer. [Copyright &y& Elsevier]

Abstract: The second Japan/US Program on Irradiation Tests for Fusion Research (JUPITER-II) began on April 1, 2001. Part of the collaborative research centers on studies of the molten salt 2LiF2–BeF2 (also known as Flibe) for fusion applications. Flibe has been proposed as a self-cooled breeder in both magnetic and inertial fusion power plant designs over the last 25 years. The key feasibility issues associated with the use of Flibe are the corrosion of structural material by the molten salt, tritium behavior and control in the molten salt blanket system, and safe handling practices and releases from Flibe during an accidental spill. These issues are all being addressed under the JUPITER-II program at the Idaho National Laboratory in the Safety and Tritium Applied Research (STAR) facility. In this paper, we review the program to date in the area of tritium/deuterium behavior, Flibe mobilization under accident conditions and testing of Be as a redox agent to control corrosion. Future activities planned through the end of the collaboration are also presented. [Copyright &y& Elsevier]

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 CaCl 2 at 1123 K. Ce-CeCl 3, In-InCl 3, and Pb-PbCl 2 were the only potential surrogates identified using these constraints. Sn-SnCl 2 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-ZnCl 2 , Sn-SnCl 2 , and Bi-BiCl 3. Zn-ZnCl 2 was used with the LiCl-CaCl 2 (65:35 mol%) eutectic at 773 K. Sn-SnCl 2 and Bi-BiCl 3 were used with the LiCl-KCl-CaCl 2 (50.5:44.2:5.3 mol%) eutectic at 673–773 K. Ce electrorefining in molten CaCl 2 resulted in a difficult to separate colloid mixture of Ce, Ca and Cl. Electrorefining rates for In in molten CaCl 2 were too slow due to InCl 3 volatilizing out of the molten salt. Only trace amounts of SnCl 2 was retained in the CaCl 2 at 1123 K resulting in impractical electrorefining rates. Zn metal product was successfully collected in the LiCl-CaCl 2 eutectic molten salt, but the metal obtained did not coalesce into one piece. Sn and Bi were successfully electrorefined in the LiCl-KCl-CaCl 2 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-SnCl 2 and Bi-BiCl 3, 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-CaCl 2 salt at 673–773 K in place of CaCl 2 at 1123 K. [ABSTRACT FROM AUTHOR]