Materials and Chemistry for Molten Salt Systems: Poster Session
Sponsored by: TMS Structural Materials Division, TMS: Corrosion and Environmental Effects Committee, TMS: Nuclear Materials Committee
Program Organizers: Stephen Raiman, University Of Michigan; Raluca Scarlat, University of California, Berkeley; Jinsuo Zhang, Virginia Polytechnic Institute and State University; Kumar Sridharan, University of Wisconsin-Madison; Nathaniel Hoyt, Argonne National Laboratory; Michael Short, Massachusetts Institute of Technology
Tuesday 5:30 PM
March 1, 2022
Room: Exhibit Hall C
Location: Anaheim Convention Center
N-9: Galvanic Corrosion in Containment Materials: Nicholas Adams1; Kerry Rippy1; Liam Witteman1; Judith Vidal1; 1NREL
While impurity mitigation and filtration can reduce the problems caused by corrosive components in molten chloride salts, the molten salt can also act as an electrolyte, causing galvanic corrosion between dissimilar containment materials. As a result, all containment metals and alloys that are in contact with the molten salt must be chosen to both resist corrosion and be galvanically compatible. Therefore, the reduction and oxidation (redox) potentials of promising corrosion-resistant alloys have been assessed. Less noble alloys, which are more vulnerable to galvanic corrosion, have been identified, and strategies to mitigate galvanic corrosion have been proposed. Techniques such as assessment of open circuit potentials (OCP) and potentiodynamic polarization sweeps were employed, and corrosion rates were calculated using the Stern-Geary equation through Faraday’s law based on the samples’ Tafel slopes.
N-10: Measurement of Vapor Pressure of UCl3 in NaCl-MgCl2-UCl3 via Transpiration Experiments: Jacob Yankey1; Easton Sadler1; Marisa Monreal2; Matt Jackson2; Scott Parker2; Suhee Choi1; Mario Gonzalez1; Michael F. Simpson1; 1University of Utah; 2Los Alamos National Laboratory
Mixtures of eutectic NaCl-MgCl2 with various concentrations of added UCl3 were used in transpiration experiments to estimate the vapor pressure and activity coefficient of UCl3 in this molten salt mixture. The motivation for this work is to develop property measurement methods relevant to molten salt reactors. A key to our approach is careful preparation of the salt. Molten NaCl-MgCl2 was thermally dehydrated and treated with anhydrous HCl gas to remove oxygen-containing species from the salt. Residual OH-/O2- concentration was measured via acid-base titration. UCl3 was then formed via chlorination of U metal with FeCl2 in-situ in the molten salt mixture and quantified via inductively coupled plasma mass spectrometry (ICP-MS). The range of U concentrations in the salt was 1-10 wt%. Vapor pressure results calculated from salt mass loss and ICP-MS analysis of salt condensate will be reported over a range of 500 to 700oC.
N-11: Methods for Recycle of Uranium in Molten Salt Reactor Fuel: Claire Perhach1; 1Caltech
For the purpose of generating UCl3 in fuel salt for a molten salt reactor (MSR), bubbling HCl into the molten salt in contact with U metal was investigated. The base salt used for this study was equimolar NaCl-CaCl2. First the NaCl-CaCl2 was purified via heating to 200°C and purified by bubbling 10% HCl gas balanced with argon. Effluent gas was fed into an auto-titrator which measured the consumption of HCl in real time and was used to determine when to stop the reaction. Then, a uranium rod (3” by 2.1mm) was inserted into salt while 10% HCl gas balanced with argon was bubbled in. Salt samples were taken and analyzed for U concentration using inductively coupled plasma mass spectrometry (ICP-MS). Applicability of this process to an integrated MSR fuel processing scheme that includes actinide drawdown from waste salt and dechlorination of waste salt will be discussed.
N-12: Molten Salt Corrosion and Irradiation Behaviors of Cladded and Surface-treated SS316H: Matthew Weinstein1; Hongliang Zhang1; Cody Falconer1; William Doniger1; Louis Bailly-Salins1; Alex Nelson1; Kumar Sridharan1; Adrien Couet1; 1University of Wisconsin-Madison
Current structural alloys, code certified for the MSRs temperature ranges, contain high levels of chromium, making them highly susceptible to salt corrosion. One potential solution to circumvent the need for code certification of new alloys is to design claddings that will protect the underlying code certified materials from corrosion damage during operation. In this work, we examined the corrosion of Ni and Cu electroplated, and carburized claddings on SS316H using static corrosion tests. The corrosion tests were performed in molten FLiNaK at 700 C up to 1000 hours. Pre- and post-corrosion Scanning Electron Microscope (SEM), energy-dispersive-spectroscopy and glow-discharge-optical-emission-spectroscopy were performed on cladding cross-sections and surfaces to evaluate degradation. Additionally, high-temperature 4MeV Ni heavy ion irradiation was performed up to 50 displacement-per-atom to assess the phase stability and irradiation behavior of the cladding systems. SEM, transmission electron microscopy and nano-indentation were performed to assess interfacial irradiation induced defects and associated hardening.
N-13: Multi-parametric Studies of Graphite Compatibility with Fluoride Salt: Miranda Mazza1; Stephen Raiman1; 1Texas A&M University
Graphite is the most commonly proposed moderator material for thermal-spectrum molten salt reactors (MSRs). Due to the porosity of graphite, salts can infiltrate into the moderate blocks, causing structural degradation and, potentially, critically hot spots. By exposing selected graphite grades to FLiNaK salt with controlled chemistry and pressure, the effect of material and salt variables can be determined. In this study, the effects of time, temperature, pore size, and pressure on fluoride salt infiltration into graphite are analyzed. This study reports on recent experiments aimed at identifying the optimal graphite grade use in the Molten Salt Research Reactor to be built at Abilene Christian University and discusses the impact of the findings for fluoride-based MSRs in general.
N-14: Process Optimization for the Purification of Molten Fluoride Salts via Gas Sparging: Kyle Williams1; Kimberly Zabava1; Stephen Raiman1; 1Texas A&M University
The physical, chemical, and nuclear properties of fluoride salt eutectics make them appealing for use in Molten Salt Reactors. Molten fluoride salts, however, can be corrosive if their redox potential is not properly managed. Sparging the salt while molten with a mixture of HF and H2 has proven to be an effective method of purification; however, little is documented about the specifics of optimizing this method. To observe the effect of various system changes on the rate and quality of salt purification, and to observe the corrosion rates and patterns of various materials in a highly fluorinating molten salt environment, a facility was constructed to purify fluoride salts based on established methods. This work reports on studies aimed at optimizing methods, chemicals, and materials used during the hydrofluorination process with an eye toward commercial scale-up.
N-15: Revealing Local Ionic Metal Structures in Molten Salt Environments Applying X-ray Absorption Spectroscopy: Luis Betancourt1; Yang Liu1; Mehmet Topsakal1; Ruchi Gakhar2; Michael Woods2; Phillip Halstenberg3; Santanu Roy4; James Wishart1; Vyacheslav Bryantsev4; Anatoly Frenkel1; Simerjeet Gill1; 1Brookhaven National Laboratory; 2Idaho National Laboratory; 3University of Tennessee Knoxville; 4Oak Ridge National Laboratory
Molten Salt Reactors (MSR) are leading candidates among Generation IV advanced nuclear power reactor designs considered for development in US. Reactivity of molten salts for the successful deployment of MSR system requires fundamental understanding of solvent chemistry, radiation induced effects as well as local structure of metals to predict their stability and reactivity. In present work, we use in-situ X-ray absorption spectroscopy (XAS) to probe the speciation of metal ions e.g transition metals (Ni,Co) as well as lanthanides (Sm,Pr) in various molten salt systems (ZnCl2-KCl, ZnCl2, LiCl-KCl) to understand the role of solvent composition as well as temperature on coordination environment of metal ion in various solvent mixtures. Evolution of local structure of various metal ions as function of solvent composition and temperature will be discussed. Work is supported by the Molten Salts in Extreme Environments, Energy Frontier Research Center, funded by the U.S. Department of Energy Office of Science.
N-16: Robust and Standardized High-temperature Molten Chloride Salt Reference Electrode: Suhee Choi1; Jim Steppan2; Michael Simpson1; 1The University of Utah; 2HiFunda LLC
Molten salts are known to be adaptable electrolytes with wide electrochemical potential windows, enabling both reduction and deposition of most metals. Therefore, molten salts have been utilized in various industries such as metal refining and pyroprocessing for recycling spent nuclear fuel. For applications, a stable reference electrode is essential to monitor and control electrochemical reactions. In this study, we investigated the stability of high-temperature molten chloride reference electrodes (MCREs) using mullite and magnesia tubes as reference electrode membrane materials in molten MgCl2-KCl-NaCl. Two electrochemical methods were used to characterize the long-term performance of MCREs at elevated temperatures (500°C): open circuit potential (OCP) versus platinum and cyclic voltammetry (CV) for up to 30 days. The OCP values and the onset potentials of chlorine evolution and magnesium reduction show similar behavior for the mullite and magnesia membranes.
N-17: Tellurium Cracking Study in Inconel 617: Ryan Gordon1; Stephen Raiman2; Lesley Frame1; 1University of Connecticut; 2Texas A&M University
Tellurium, a byproduct of fission in molten salt reactors, has been shown to cause or amplify cracking in Alloy N, a low-Cr Ni-based alloy. Limited research has shown tellurium impacts mechanical properties and may cause embrittlement in 316 stainless steel and other nickel alloys. Inconel 617 is a nickel-based alloy which has just recently been approved for use in nuclear environments. Tensile tests will be conducted to determine if tellurium also causes embrittlement and impacts the mechanical properties of Inconel 617. Tensile samples were heated in a tellurium rich atmosphere and strained until failure. The microstructures were analyzed, and stress strain curves were created to compare the mechanical properties of samples exposed to tellurium with samples that were not. This research shows the initial findings to determine whether tellurium-based cracking is a degradation mode of concern in molten salt reactors.
Cancelled
N-18: The Reduction of Uranium Dioxide Pellet in Molten CaCl2-CaF2-CaO: Nagihan Karakaya1; Jinsuo Zhang1; 1Virginia Tech
Reprocessing the used nuclear fuel has excessive attention due to reducing the radiotoxicity and waste generation all around the world. Direct electrochemical reduction of the fuel pellet can present an influential pathway to obtain uranium metal from oxide fuel pellet; however, to achieve highly efficient recovery of uranium, the salt bath choice has an important role such as reduction temperature that depends on melting temperature of salt, impurity products that may affect reduction reaction. In this study, the UO2 reduction is carried out in CaCl2 - CaF2 -CaO salt composition at 750C, 850C, and 950C. Oxygen removal from the system is conducted producing CO and CO2 by using chronopotentiometry and cyclic voltammetry is used to check impurity production such as carbonate in the system. Outputs are characterized by XRD and SEM-EDS analyses.