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A method of ranking the rate of corrosion of metals in molten fluoride or chloride salts is proposed based on a zero-resistance ammeter (ZRA). The metal of interest for the corrosion study is shorted to a relatively noble metal counter electrode. Stainless steel 316 was tested in FLiNaK (LiF-NaF-KF). Haynes 230, stainless steel 316L, and Hastelloy C-22 were tested in MgCl2-NaCl-KCl. The ZRA oxidation current from SS-316L was reduced by 89% in FLiNaK after the addition of Li metal, which is known to reduce the redox potential. Post-test examination of metal surfaces and salt samples were also consistent with the relative ZRA current response in fluoride salt. Consistency was also observed between the relative ZRA response for the different metal alloys in the chloride salt and post-test analysis of metal and salt samples.

CaCl2 has applications for electrochemical processing of nuclear materials. Thermal dehydration leads to formation of oxide ions, which are shown to react and cause precipitation of dissolved CeCl3 that was selected as a surrogate for actinide chlorides. Thus, measurement of residual water in CaCl2 is an essential capability. Thermogravimetric analysis (TGA) was shown to underestimate starting water concentration. Subsequent analysis of solid samples via acid–base titration and cyclic voltammetry (CV) of the molten salt yielded consistent values within 5% for residual water. Hydroxides were shown to be unstable, thus oxygen is retained as oxide ions. Thus, total water in a sample of CaCl2 can be quantified by combining TGA with with either CV of molten salt or titration of salt samples. The importance of quantifying residual water in the salt was demonstrated by showing that cerium chloride (surrogate for actinide chlorides) will react with oxide ions in CaCl2 to form insoluble oxides and oxychlorides.

4H-SiC alpha detectors were fabricated with a 21-μm thick depletion depth and were packaged into a stainless-steel casing with a mineral insulation cable and a standard BNC connector. The packaged detectors had a resolution of 0.624% FWHM at 5.486 MeV prior to salt immersion. The detectors were then immersed in a LiCl–KCl–UCl3 molten salt at 500 °C, from which a thin layer of depleted uranium was electrodeposited onto the detectors. Alpha particle emission spectra were collected from the electrodeposited source. The energy resolution of the surviving detector was 2.29% FWHM at 4.198 MeV and was sufficient to separate the 234U from 238U alpha emissions (577 keV difference). The 234U/238U activity ratio and the isotopic concentrations of 234U and 238U were determined and are representative of the uranium source used in the electrodeposition.

Dechlorination of eutectic LiCl–KCl based electrorefiner (ER) salt is reported via ion-exchange reaction with protonated ultrastable Y-type (USHY) zeolite bound into mechanically fluidized 45–250 μm diameter particles. Evidence of exchange of cations from the salt (Li+, K+, and fission product cations) into the zeolite lattice replacing H+ ions was found based on a change in unit cell size, ICP-MS, XRD and TEM–EDS in addition to detection of HCl off gas. Ion exchange reaction was carried out at 625 and 650 °C, temperatures above the melting point of eutectic LiCl–KCl. Experiments were carried out to optimize zeolite drying temperature, estimate maximum ion-exchange capacity, and determine the thermal stability of USHY zeolite. The results indicate over 90% dechlorination can be achieved without zeolite structure collapse at 625 °C. This provides a promising route to stabilizing waste from radioactive chloride salts into dechlorinated waste forms for permanent geologic disposal.

The alpha emission spectrum of electrodeposited thin film 232Th and depleted uranium sources, including the alpha emitting 232Th decay chain daughters and 238U daughter nuclide 234U, have been measured using a thin film 4H–SiC alpha detector. The energy resolution of this detector was proved suitable for the identification of many of the major characteristic alpha emissions associated with the 232Th daughter isotopes. The identification of the lower energy daughter isotopes made from the 4H–SiC detector measurements was confirmed using a silicon surface barrier detector. A depleted uranium source was fabricated using electrodeposition in a molten salt and the alpha spectrum was acquired with a 4H–SiC detector. The 234U/238U activity ratio and atomic enrichment of 234U in the sample were determined. A method of determining the 235U enrichment in the product stream of the enrichment process used to produce the depleted uranium source was developed. This method makes use of the 234U and 235U waste stream enrichments, as indicated by the depleted uranium source, and a mass balance equation to determine the product 235U enrichment. Since the 235U spectrum was obscured in the electrodeposited source by peak broadening of 238U and 234U, a table of common 235U waste enrichment values was used to compute 235U product enrichments, thereby resulting in a plausible identification of the enrichment process as a Department of Energy HEU enrichment process.

Alpha spectroscopy can be used to quantify actinide (e.g., U, Pu) concentration in a molten salt electrorefining environment. One could electroplate actinide samples directly onto a semiconductor alpha particle detector to obtain representative isotopic concentrations from a measured alpha particle energy spectrum. In this work, we fabricated a SiC Schottky device that can be partially biased to a depletion depth of 8.8 µm that is able to measure the energy spectrum of 4.012 MeV alpha particles emitted from a thorium film with a thickness. We also present a method in calculating the thickness of a medium thick alpha source with a dE/dx detector.

The solubility of lithium metal in molten LiCl–Li2O mixtures has been measured at various concentrations of Li2O ranging from 0 wt.% to 2.7 wt.% at a temperature of approximately 670–680°C. After contacting molten lithium with molten LiCl–Li2O for several hours to achieve equilibrium saturation, samples were taken by freezing the salt onto a room-temperature steel rod and dissolving in water for analysis. Both volume of hydrogen gas generated and volume of titrated HCl were measured to investigate two different approaches to calculating the lithium concentration. There appeared to be no effect of Li2O concentration on the Li solubility in the salt. But the results vary between different methods of deducing the amount of dissolved Li. The H2 collection method is recommended, but care must be taken to ensure all of the H2 has been included.

A ceramic waste form is being developed by Argonne National Laboratory* (ANL) as part of the demonstration of the electrometallurgical treatment of spent nuclear fuel [1]. The halide, alkaline earth, alkali, transuranic, and rare earth fission products are stabilized in zeolite which is combined with glass and processed in a hot isostatic press (HIP) to form a ceramic composite. The mineral sodalite is formed in the HIP from the zeolite precursor. The process, from starting materials to final product, is relatively simple. An overview of the processing operations is given. The metrics that have been developed to measure the success or completion of processing operations are developed and discussed. The impact of variability in processing metrics on the durability of the final product is presented. The process is demonstrated to be robust for the type and range of operation metrics considered and the performance metric (PCT durability test) against which the operation metrics are evaluated.