Synergistic Irradiation, Corrosion, and Microstructural Evolution in Nuclear Materials: Irradiation-Corrosion of Materials in Light Water Reactors I
Sponsored by: TMS Structural Materials Division, TMS: Nuclear Materials Committee
Program Organizers: Djamel Kaoumi, North Carolina State University; Michael Short, Massachusetts Institute of Technology; Peter Hosemann, University of California, Berkeley; Stephen Raiman, University of Michigan; Raluca Scarlat, University of California, Berkeley; Aaron Kohnert, Los Alamos National Laboratory; Ryan Schoell, Sandia National Laboratory; Philip Edmondson, The University of Manchester; Celine Cabet, Commissariat a l'Energie Atomique

Tuesday 8:00 AM
March 1, 2022
Room: 202A
Location: Anaheim Convention Center

8:00 AM  Invited
Synergistic Processes in Irradiation-corrosion of Materials in High Temperature Water: Rigel Hanbury1; Peng Wang1; Stephen Raiman1; Gary Was1; 1University of Michigan
    The reactor core is a multi-component extreme environment in which materials must operate. When combined, various stressors on core materials (temperature, stress, chemical, irradiation) produce degradation rates and modes that are not simply additive nor predictable. The effects of simultaneous irradiation and corrosion are also not separable (e.g. post-irradiation corrosion tests) and also provide an excellent example of the variability in material response with environment. For example, in high temperature water, corrosion of Zircaloy is greatly accelerated under irradiation with corrosion rates increasing steeply with flux, while corrosion of stainless steel under irradiation is suppressed. Corrosion of SiC depends strongly on the fabrication process and on both the corrosion potential and irradiation. The nature of irradiation-corrosion interaction in various systems will be discussed along with proposed processes to explain some of the seemingly contradictory observations.

8:30 AM  
Effect of Ion Irradiation on the Corrosion of 304SS in PWR Simulated Water Chemistry: Fu-Yun Tsai1; Ryan Schoell1; Khalid Hattar2; Djamel Kaoumi1; 1North Carolina State University; 2Sandia National Laboratories
    A study of the effects of ion irradiated microstructure on corrosion of 304SS in simulated Pressurized Water nuclear Reactor conditions was conducted. Coupons of 304SS were irradiated with 10 MeV Au ions to 10 displacements per atom (averaged over 1 μm thickness below surface). Corrosion was conducted in a recirculating autoclave in simulated PWR primary water condition (325 ℃, 2200 psi, 30 cc H2/kg of H2O at STP) for 1, 8,14, 21 and 35 days and 63 days. Transmission Electron Microscopy (TEM) of cross-sectional samples was used to characterize the oxide and its growth kinetics through oxide-thickness measurements in both irradiated and non-irradiated regions. Results showed differences in oxide thickness and shape between irradiated and non-irradiated regions although the oxide phases were similar. A discussion was conducted to substantiate the effects of irradiation.

8:50 AM  
Investigation of Hydrogen Trapping by Irradiation-induced Defects in 316L Stainless Steel: Frantz Martin1; Anne-Cécile Bach2; S. Perrin3; F. Jomard4; Cecilie Duhamel4; Jerome Crepin5; 1Université Paris-Saclay, CEA; 2Université Paris-Saclay, CEA, MINES ParisTech, PSL Research University; 3CEA, DES, ISEC, DE2D, SEAD, LCBC, University of Montpellier; 4Groupe d'Etude de la Matière Condensée, CNRS, UVSQ; 5MINES ParisTech, PSL Research University
     In-core stainless steel components of nuclear power plants undergo both corrosion and neutron bombardment: the first provides a continuous hydrogen flux into the material while the latter induces structural damages creating point and/or extended defects. Structural integrity assessment must consider these interactions between hydrogen and such defects. 316L stainless steel was irradiated with two Fe-ion beams chosen to create irradiation defects close those encountered in neutron-irradiated 316L SS in terms of density and size. Dislocations loops and nanocavities densities, sizes and distributions were characterized by TEM. The specimens were then cathodically charged in deuterium and analyzed by SIMS. It highlighted that radiation-defects can trap deuterium, mainly at nanocavities. It was also shown that deuterium remained trapped even during heavy primary water exposure. The results will be discussed in terms of trapping/detrapping kinetic constants considering the internal hydrogen pressure of nanocavities and the dissociation energy.

9:10 AM  
Effects of Water Radiolysis and Displacement Damage during Simultaneous Irradiation and Corrosion of 316L Stainless Steel: Rigel Hanbury1; Jonas Heuer2; Gary Was1; 1University of Michigan; 2Naval Nuclear Laboratory
    Proton irradiation of 316L stainless steel discs was conducted during exposure in both water and steam environments for 24 h and 72 h with a damage rate of 7×10-7 dpa/s. To isolate radiolysis with minimal displacement damage, additional 316L sample surfaces oriented parallel to the transmitted proton beam were examined. To isolate displacement damage, a steam environment was used to minimize water radiolysis without changing the damage rate. Relative to a non-irradiated control, total oxidation was decreased by radiolysis and unaffected by displacement damage in 320 °C hydrogenated water. In 480 °C hydrogenated argon-steam, total oxidation was unaffected by radiolysis and increased with displacement damage relative to the non-irradiated control. The high level of water radiolysis by ion irradiation suppressed corrosion by increasing the electrochemical corrosion potential. In steam, where radiolysis was suppressed, displacement damage accelerated oxidation consistent with a transport-limited corrosion process.

9:30 AM Break

9:50 AM  
Effects of Pre-irradiation on the Corrosion Behavior of I600 in Hydrogenated Water: Ryan Schoell1; Fu-Yun Tsai1; Peter Baldo2; Yongqiang Wang3; Djamel Kaoumi1; 1North Carolina State University; 2Argonne National Laboratory ; 3Los Alamos National Laboratory
    A study was conducted to characterize the effects of pre-irradiated microstructure on the corrosion behavior of I600 in hydrogenated water at 2200 psi and 325 °C (representative of Pressurized Water Reactor conditions). Coupons were polished to a mirror finish and partially irradiated with either 1 MeV Ar ions or 5 MeV Ni ions to 10 dpa (as averaged over the first 100 nms). The samples were corroded in the simulated PWR environment inside a recirculating autoclave. The hydrogen level was kept around 30 cc/kg H2O at STP by continuously bubbling in an overpressure of hydrogen. Transmission Electron Microscopy (TEM) characterization was conducted after testing for several durations (7 days, 14 days, 28 days). The oxide thicknesses were measured and the oxide microstructure was characterized and compared in the irradiated vs. unirradiated areas. The effect of the irradiated microstructure on oxide growth kinetics and mechanisms is discussed.

10:10 AM  Cancelled
NOW ON-DEMAND ONLY - The Role of Surface and Interfacial Chemistry in Hydrogen Corrosion of Uranium: Shohini Sen-Britain1; Yaakov Idell1; Wigbert Siekhaus1; Kerri Blobaum1; Bill Mclean1; 1Lawrence Livermore National Laboratory
    Hydrogen corrosion of uranium results in the formation of uranium hydride (UH3) precipitates that mechanically deform the surface. Here, we investigated the surface chemistry of complete and partially erupted regions of a depleted uranium (DU) surface after H2 gas exposure using ion microscopy, focused ion beam (FIB) cross-sectioning, and transmission electron microscopy (TEM). This multimodal characterization provided novel insight into hydrogen corrosion of DU. We report that partial and complete surface eruption occurs at or near carbide and aluminide inclusions at the DU surface. Erupted regions contain a layer of uranium hydride enriched at the surface; meanwhile, the partially erupted regions contain uranium hydride buried below the surface. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under contract DE-AC52-07NA27344.

10:30 AM  Cancelled
Cooperative Role of Pressure- and Corrosion on Stress Buildup of the High Temperature LWR Pipes: Asghar Aryanfar1; Jaime Marian2; 1American University of Beirut; 2University of California Los Angeles
    The corrosion in the pressurized cooling pipeline of LWR is a catastrophic event leading to the fracture. We develop a real-time framework for the accumulation of compressive stress due to both elastic stress from the imposed internal/external pressure and the corrosion stress from the hosting oxygen in the metal matrix. In this regard, we quantify the infiltration of the oxygen within the metal of the curved boundary via predicting the proper space-time segmentation during the high-temperature exposure and we compute the rate of the oxide growth stoichiometrically. Consequently, the onset of the failure is predicted in real-time and based on the amount of the accumulated stresses.