Synergistic Irradiation, Corrosion, and Microstructural Evolution in Nuclear Materials: Irradiation-Corrosion of Materials in Light Water Reactors II
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 2:30 PM
March 1, 2022
Room: 202A
Location: Anaheim Convention Center
2:30 PM Invited
Decoupling of Ion Irradiation Effects on the Corrosion Rate of Zirconium Alloys: Experiment and Modelling: Marc Tupin1; 1CEA
Understanding the corrosion processes of fuel rod claddings is still a challenge regarding impact of alloying elements on the oxidation rate. The corrosion mechanisms of zirconium alloys - Zircaloy-4 and M5 - are not fully understood yet, owing to the complexity of these materials, the environment and the ionizing radiation conditions. The goal of our last decade’s research was to decorrelate and separately study the influence of radiation damage into the metal from the induced effects within the oxide layer, the water radiolysis impact. The most remarkable finding is that the radiation effects on the corrosion rate of Zircaloy-4 are broadly speaking negative, whereas M5 alloy, exhibits a better behaviour after ion irradiation. We also studied the relationship between the oxidation process and the H-PickUp phenomenon, and the effect of ion irradiation on the hydrogen pickup fraction. Results and prospects will be discussed in terms of mechanism and modelling.
3:00 PM Cancelled
Modeling Zircaloy Oxidation and Hydridation under Irradiation: What Can be Learned from Physical Models and Simulations: Jaime Marian1; 1University of California, Los Angeles
Understanding the corrosion behavior of zirconium alloys used in nuclear reactors remains a challenging undertaking due to their microstructural complexity and the multiparametric nature of operation conditions. Modeling and simulation can provide a useful testbed on which to verify and validate physical hypothesis and gain insight into complex chemo-physical behaviors by isolating a set of selected operational variables and study their effect on material performance. In this work, we develop a multilayer model of oxide scale growth in Zr and Zr alloys, incorporating several physical mechanisms to explain the observed oxidation kinetics in these materials. Furthermore, we connect the models to measurements obtained during dedicated irradiation experiments performed at the Michigan Ion Beam Laboratory in an attempt to separate the effects of irradiation on corrosion kinetics. Finally, we present modeling results of incipient hydride cluster formation during dynamic oxidation conditions.
3:30 PM
Ion Irradiation Study of Polymer Derived SiFeOC-C-SiC Composite: Kathy Lu1; Sanjay Kumar Singh1; 1Virginia Polytechnic Institute and State University
Silicon oxycarbides (SiOCs) are known for their resistance to crystallization, oxidation, and creep at high temperatures. There is great potential for the development of uniform C-SiOC coatings for TRISO fuels in advanced nuclear reactors. Use of iron precursor in SiOC not only helps the formation of nanodomains of β-SiC and FeSi but also enhances the mechanical integrity of the C-SiOC coatings. In this study, SiFeOC-C-SiC ceramics are prepared using a polymer derived method and irradiated using 400 keV Kr ions to 10 and 50 dpa at room temperature and 600°C. The effect of nanodomain interfaces (graphite, SiC, Fe and Fe3Si) in acting as a point defect sinks to mitigate the radiation tolerance and swelling are investigated. Oscillation of free volume or local bonding arrangement to recover itself by self-healing mechanism is elucidated. The effects of ion irradiation on the mechanical properties are studied using nano-indentation.
3:50 PM Invited
Dramatic E-beam Enhancement of Zircalloy-4 Corrosion: David Bartels1; 1Notre Dame University
Experimental results will be presented for remarkable enhancement of zircalloy-4 corrosion by 2.5MeV E-beam (0.16mA/cm2) exposure. The 400 micron thick samples form the interface between accelerator vacuum, and deionized, H2-saturated water held at high temperature under 150atm pressure. Experiments were conducted both as a function of exposure time and at several temperatures. Relative to thermal corrosion rates, the E-beam acceleration is startling. At 100oC, an oxide layer of 0.8 microns was generated in only 12 hours. This would take decades in an autoclave.Acceleration of this magnitude can only arise if the E-beam directly affects the rate-limiting step. Corrosion requires both O- ion transport and electron transport through the growing zirconium oxide layer; either process could be rate-limiting. We believe this E-beam experiment provides definitive proof that electron transport is rate-limiting in the initial stages of zircalloy corrosion, a conclusion that runs counter to most models in the literature.
4:20 PM Break
4:40 PM
Understanding the Hydrogen Pickup Mechanism of M5Framatome under Ion Irradiation: Benoit Queylat1; Michael Jublot1; Frantz Martin1; Francois Jomard2; Marc Tupin1; 1Commissariat a l'Energie Atomique et aux Energies Alternatives (CEA); 2Universite de Versailles St-Quentin-en-Yvelines
While Zircaloy-4 alloy shows a high and constant hydrogen pickup fraction (HPUF) over its operating life in Pressurized Water Reactor (PWR), this fraction is lower for M5Framatome alloy and tends to decrease for high burnup. One of the hypotheses to explain such a difference is the positive role of the continuous neutron irradiation during corrosion. The role of irradiation on the HPUF in post transition was simulated with ion irradiation of pre-oxidized M5Framatome samples. Oxidation and hydrogen pick up kinetics were monitored for 50 days during post irradiation corrosion tests in PWR conditions in static autoclave. Isotopic exposure in heavy water combined with Secondary Ion Mass Spectrometry (SIMS) were used to study the impact of irradiation on hydrogen diffusion in the oxide layer. A hydrogen pickup mechanism and a model of the HPUF evolution of M5Framatome claddings have been suggested.
5:00 PM Invited
Electrochemical Behavior of Fuel Cladding under In-situ UV Irradiation: Adrien Couet1; Taeho Kim1; Antoine Ambard2; 1University of Wisconsin-Madison; 2Electricite de France
The effect of Cerenkov (i.e. UV) radiation on the electrochemical response of Zircaloy-4 fuel cladding material in reducing high-temperature water conditions is investigated with open circuit potential (OCP), electrochemical impedance spectroscopy and Mott-Schottky analysis. At room temperature, significant OCP drop is observed when UV is irradiated on pre-oxidized samples, representing n-type oxide behavior. The oxide impedance of pre-oxidized Zircaloy-4 decreases under UV irradiation, likely due to a reduction of the oxide capacitance. Under in-situ conditions (320 ℃ and 15 MPa), the flat-band potential and donor density increase under UV irradiation, confirming electron-hole pair generation and additional band bending at the oxide/water interface. SEM and TEM characterizations clearly show magnetite crystals nucleate and grow on top of zirconium oxide under UV irradiation, while none are observed in non-irradiated area. A UV irradiation-induced photo-electrochemical mechanism involving Fe photo-deposition and oxide photo-dissolution and its implication on in-reactor fuel cladding performance are discussed.
5:30 PM
Atomic Level Mechanisms Underlying Hydrothermal Corrosion of Silicon Carbide in Nuclear Reactors: Jianqi Xi1; Dane Morgan1; Izabela Szlufarska1; 1University of Wisconsin-Madison
Silicon carbide (SiC) has been recognized for its potential as the structural material in the accident tolerant fuel in the advanced light water reactors. However, one of the remaining issues that can limit its application is the hydrothermal corrosion. In this talk, I will discuss our recent theoretical studies revealing corrosion mechanisms of SiC in both the region of crystalline and general grain boundaries (GBs) as exposed to the hydrothermal environment. I will first discuss the elementary interfacial reactions driving corrosion on the crystalline surfaces, including a discovery of the unexpected hydrogen scission reaction that plays a key role in surface degradation. Our kinetic studies reveal that SiC is dissolved directly into the water without forming the silica layer, although the reactions are analogous to those observed during dissolution of silica. Secondly, I will discuss the influence of irradiation-induced segregation at the general GBs on the hydrothermal corrosion of SiC.