Materials in Nuclear Energy Systems (MiNES) 2021: Integrated Phenomena- Session I
Program Organizers: Todd Allen, University of Michigan; Clarissa Yablinsky, Los Alamos National Laboratory; Anne Campbell, Oak Ridge National Laboratory
Wednesday 8:00 AM
November 10, 2021
Room: Allegheny
Location: Omni William Penn Hotel
Session Chair: Jesse Carter, Naval Nuclear Laboratory
8:00 AM Invited
Radiation-decelerated Corrosion of Nuclear Structural Materials in Gen IV Reactor Environments: Weiyue Zhou1; Nouf AlMousa1; Kevin Woller1; Guiqiu Zheng1; Yang Yang2; Michael Lastovich3; Ryan Schoell3; Peter Stahle1; Angus Wilkinson4; Michael Moody4; Andrew Minor2; Tom Lapington4; Felix Hofmann4; Djamel Kaoumi3; Michael Short1; 1Massachusetts Institute of Technology; 2Lawrence Berkeley National Laboratory; 3North Carolina State University; 4Oxford University
The effects of ionizing radiation on materials often reduce to “bad news.” Radiation damage usually leads to detrimental effects, including radiation-accelerated corrosion. However, we have discovered a subset of conditions, practically useful for nuclear structural materials, where radiation damage decelerates corrosion. For example, proton irradiation decelerates intergranular corrosion of Ni-Cr model alloys and Ni-rich commercial alloys in molten fluoride salt at 600-700°C. We demonstrate this by showing that the depth of intergranular voids resulting from Cr leaching into the salt is reduced by proton irradiation alone. Radiation enhanced diffusion more rapidly replenishes corrosion-injected vacancies with alloy constituents, playing the crucial role in decelerating corrosion. Analogous results for steels in molten lead will also be shown in this talk. Only such fully coupled experiments can show that irradiation can have a positive impact on materials performance, challenging our view that radiation damage always results in negative effects.
8:40 AM
Mitigating Irradiated Assisted Stress Corrosion Cracking with Minor Refractory Element Modification – A High-throughput Approach Using Compositionally-graded Specimen: Jingfan Yang1; Laura Boring2; Lingfeng He3; Miao Song4; Zhijie Jiao4; Yongfeng Zhang5; Daniel Schwen3; Lin Shao2; Xiaoyuan Lou1; 1Auburn University; 2Texas A&M Univeristy; 3Idaho National Laboratory; 4University of Michigan; 5University of Wisconsin
Irradiation experiments and post-irradiation material testing are critical for the new alloy development and qualification in reactor cores. It represents the most costly and lengthy step, and typically follows a one-at-a-time experimental paradigm. This study demonstrates the feasibility of using compositionally gradient specimens, fabricated by laser additive manufacturing, to improve the throughput of stainless-steel composition screening for better irradiated assisted stress corrosion cracking (IASCC) resistance. The low-level doping (<1 wt.%) of minor refractory elements was selected because of its significance on the local atomic level heterogeneity. The study confirmed these elements not only interacted with point defects but also resulted in different grain boundary oxidation, as well as grain-boundary-dislocation interaction. The validity of using compositionally-graded specimens in IASCC evaluation has been demonstrated for two testing methods, constant extension rate test and step load test. The advantages and challenges of using compositionally gradient design for IASCC evaluation will be discussed.
9:00 AM
Determination of Tritium Trapping Mechanisms in the TPBAR Aluminide Coating: Anne Chaka1; Bruce Schmitt1; 1Pacific Northwest National Laboratory
The 316SS cladding and its iron aluminide coating of the Tritium Producing Burnable Absorber Rods (TPBAR) serve as both a structural pressure boundary and a permeation barrier in order to retain tritium within the TPBAR. Experimental results have determined that tritium is retained in the aluminide barrier region at a concentration several orders of magnitude greater than can be explained by solubility alone, suggesting that other trapping or binding mechanisms are involved. The cladding is subject to both hydrogen and tritium flux, so the binding energies of both need to be determined. We utilize first principles quantum mechanics (density functional theory) in combination with ab initio thermodynamics to determine the free energy of binding for tritium and protium in the aluminum-rich phases of the aluminide coating as a function of temperature and pressure. Both Fe and Al vacancy sites are evaluated as well as interstitial sites.
9:20 AM
Understanding Tritium Permeation in FeCrAl Alloys: Andrew Hoffman1; Kan Sakamoto2; Yogen Garud3; Xunxiang Hu4; Fabiola Cappia5; Raul Rebak1; Rajnikant Umretiya1; 1GE Research; 2Nippon Nuclear Fuel Development Co.; 3SIMRAND, LLC; 4Oak Ridge National Laboratory; 5Idaho National Laboratory
FeCrAl alloys are an excellent candidate for accident tolerant nuclear fuel cladding due to their corrosion resistance and good mechanical properties. One potential topic for concern recently has been the hydrogen (and in particular tritium) permeation behavior of these alloys. Because ferritic alloys have the potential for having higher hydrogen permeation the current commercial Zr based alloys, tritium generated in the fuel could be released into the coolant water. This presentation will give an overview of the current work being done to understand permeation behavior in ferritic FeCrAl alloys. Attention will be given to the oxide layers which can be generated from the fuel-cladding interface and the coolant cladding interface. Such oxide layers are presumed to significantly decrease the hydrogen permeation in FeCrAl alloys.
9:40 AM Break