Advanced Characterization of Materials for Nuclear, Radiation, and Extreme Environments: Materials Mechanics & Environmental Degradation
Sponsored by: TMS Nuclear Materials Committee
Program Organizers: Samuel Briggs, Oregon State University; Christopher Barr, Department Of Energy; Emily Aradi, University of Huddersfield; Michael Short, Massachusetts Institute of Technology; Janelle Wharry, Purdue University; Cheng Sun, Clemson University; Dong Liu, University of Oxford; Khalid Hattar, University of Tennessee Knoxville
Wednesday 2:00 PM
November 4, 2020
Room: Virtual Meeting Room 13
Location: MS&T Virtual
Session Chair: Michael Short, Massachusetts Institute of Technology; Samuel Briggs, Oregon State University
2:00 PM Invited
Explaining the Corrosion Morphology of Structural Materials in Molten Fluoride Salts With/Without Radiation: Weiyue Zhou1; Yang Yang2; Miaomiao Jin3; Lingfeng He3; Andrew Minor2; Michael Short1; 1Massachusetts Institute of Technology; 2Lawrence Berkeley National Laboratory; 3Idaho National Laboratory
Corrosion of Cr-bearing structural metals in molten salts proceeds mainly via selective leaching of Cr. Apart from understanding the interfacial reaction and its thermodynamics, the kinetics of mass transport within the alloys are equally important. One of the biggest challenges is interpreting the observed corrosion morphology. By exploring the parameter space of temperature and concentration, we categorize the morphology of the response into several groups. Advanced characterization is critical to elucidating each mechanism. For MSR applications, it is also crucial to understand the synergy effect of radiation and corrosion. By running simultaneous proton irradiation and corrosion experiments on Ni-Cr model alloys, we show that proton irradiation, under certain conditions, can decelerate corrosion in molten salts. Armed with knowledge of the corrosion mechanisms responsible, we will present the kinetics behind the observed deceleration to enable prediction of its utility in MSRs.
2:40 PM
Enabling In-situ Crack Growth Testing and Monitoring in VTR Cartridge Loop Environments: Samuel Briggs1; Peter Beck1; Dustin Mangus1; Jake Quincey1; Andrew Brittan1; George Young2; Guillaume Mignot1; Julie Tucker1; 1Oregon State University; 2Kairos Power
The Versatile Test Reactor (VTR) is a proposed fast-spectrum research reactor being developed by the U.S. Department of Energy to aid in design and licensing of next-generation nuclear reactors. While the primary coolant will be liquid sodium, the proposed design incorporates self-contained cartridge loops, enabling experimentation in other advanced reactor coolant environments, such as molten salts, gases, or other liquid metals. Efforts at Oregon State University are focused on developing techniques enabling fully-instrumented in-situ environmentally-assisted cracking (EAC) experiments in various cartridge loop environments. To date, EAC test facilities capable of corrosion fatigue, stress corrosion cracking, and liquid metal embrittlement studies in liquid sodium, molten salt, and supercritical CO2 environments have been developed. In addition, various non-destructive testing techniques, including potential drop and acoustic emission monitoring, are being adapted for use in cartridge loop environments and geometries. A general overview of these efforts and initial results will be presented.
3:00 PM
Unveiling High Temperature Damage Mechanisms via In-situ Digital Image Correlation of Chromium-coated Zirconium-based Fuel Claddings: David Roache1; Alex Jarama1; Clifton Bumgardner1; Frederick Heim1; Morgan Price1; Xiaodong Li1; 1University of Virginia
Coated nuclear fuel claddings offer a promising, near-term solution to address the demand for next-generation, accident-tolerant fuel systems and possess superior mechanical properties and greater oxidation resistance compared to current cladding technology, allowing for improved performance during beyond design-basis accident conditions. Here, we unveil the high temperature failure mechanisms of chromium-coated zirconium alloys at temperatures up to 1200°C using a novel mechanical test rig coupled with in-situ three-dimensional digital image correlation and acoustic emissions sensing to monitor spatial strain and crack initiation / propagation during cladding expansion. Ex-situ optical and scanning electron microscopy were used to characterize crack propagation at various levels of strain and temperature, and a 2D fracture model was created to assess the effects of temperature and crack size on cladding fracture energy. We observed evolving fracture mechanisms beginning at temperatures as low as 300 °C, which will carry significant implications for their use in reactor environments.
3:20 PM
Development of an In-Situ Mechanical Test System for Advanced Reactor Coolants: Jake Quincey1; Peter Beck1; Josef Parrington2; Lars Parrington2; Christopher Lamb2; Henry Korellis3; Pit Schulze3; Alan Kruizenga3; Micah Hackett3; George Young3; Julie Tucker1; Samuel Briggs1; 1Oregon State University; 2Parrington Instruments; 3Kairos Power
Advanced reactor coolants such as liquid sodium and molten salts pose unique obstacles for test systems designed to study environmentally assisted cracking (EAC). Challenges with in-situ testing include high temperature operation with simultaneous coolant chemistry control, all while necessitating access to the sample for real-time load control and monitoring of crack initiation and growth. The present work has developed an in-situ mechanical test system that addresses these issues and is capable of state-of-the-art, fracture mechanics-based EAC testing in liquid sodium and molten salt coolant environments. This test system is constructed from stainless steel to minimize dissimilar metal contact, enables electrical isolation of the test sample for potential drop monitoring of cracking, and utilizes a novel pumped ‘cold leg’ to prevent excessive corrosion product buildup in the coolant. The key features of the system and initial results from testing in FLiNaK will be described.
3:40 PM
Corrosion Control of Austenitic Stainless Steel and Nickel-Based Alloys in Molten Chloride Salt Environments: Kasey Hanson1; Krishna Moorthi Sankar1; Remi Dingreville2; Joshua Sugar2; Chaitanya Deo1; Preet Singh1; 1Georgia Institute of Technology; 2Sandia National Laboratories
Chloride salts have been studied for use in nuclear Molten Salt Reactors (MSRs). MSRs offer a viable alternative to light-water reactors, largely due to their ability to operate at low-pressures and superior heat-transfer properties. Thermodynamically driven selective dissolution of alloying elements due to impurities present within the eutectic chloride salt mixture have been identified to be the biggest contributor to material degradation. Research in identifying materials suitable for this environment have largely centered on nickel-based alloys, due to the alloy’s stability resulting from minimal nickel dissolution into the eutectic salt mixture. This research tested nickel-based alloys and one austenitic stainless steel in a KCl—MgCl2 eutectic mixture at 700°C with redox control of the eutectic chloride salt as a method for reducing material degradation and studied what impact such a method may have on material properties of exposed alloys.
4:00 PM
Design of a Hot Hydrogen Test Loop for Testing of Nuclear Thermal Rocket Elements: William Searight1; Alex Somers1; Leigh Winfrey1; 1The Pennsylvania State University
Nuclear thermal propulsion (NTP) is a promising candidate for deep-space manned missions to Mars for its powerful, compact engines, which offer transit times approximately half that of traditional chemical rockets. One of the most pressing issues in the development of NTP rocket engines is the design and testing of high temperature materials which can withstand the high heat and particle fluxes from fission in the core and maintain desirable surface and inner molecular structure and material performance. To this end, a hot hydrogen test loop capable of producing circulating hydrogen plasma at temperatures up to 3000 K is being designed and constructed at Penn State to study the plasma-material interactions of plasmas with NTP component materials. This work focuses on the ongoing progress in modeling and designing this tabletop-sized loop to test tie tubes, U-bend shaped tubes through which the hydrogen coolant/propellant will be channeled through the NTP core.
4:20 PM
Development of a Combined Thermal Hydraulic and Materials Corrosion Liquid-Sodium Experimental Facility: Dustin Mangus1; Juwan Johnson1; Brett Leitherer1; Peter Beck1; Seth Walton1; Guillaume Mignot1; Wade Marcum1; Julie Tucker1; Samuel Briggs1; 1Oregon State University
The Versatile Test Reactor (VTR) program seeks to design and construct a fast-spectrum research reactor to assist in addressing the capability gap in the testing of advanced core materials, fuels, and instrumentation in prototypical next-generation nuclear reactor environments. Current design concepts utilize the proven experience of pool-type sodium-cooled reactors. In support of this program, Oregon State University has developed the Glovebox for Experimental Liquid Sodium (GELS) facility to support the development of test rigs and instrumentation enabling environmentally assisted cracking experiments with in-situ monitoring capabilities in proposed cartridge loop environments. This facility allows for chemistry-controlled flowing or static liquid sodium test capabilities, facilitating both thermal-hydraulic and materials corrosion experimental needs. This is done using a secondary diagnostics loop with oxygen cold-trapping capabilities and a combination of conduction and moving-magnet pumps. An overview of facility design specifications and preliminary experimental results will be presented.