Corrosion behavior of additively manufactured FeCrAl in out-of-pile light water reactor environments
Not Available
Materials Science↗
Engineering topics
Rebak, Raul B.
Publications and source records attributed to Rebak, Raul B..
Hydrothermal Corrosion of Latest Generation of FeCrAl Alloys for Nuclear Fuel Cladding
After the Fukushima nuclear disaster, the nuclear materials community has been vastly investing in accident tolerant fuel (ATF) concepts to modify/replace Zircaloy cladding material. Iron–chromium–aluminum (FeCrAl) alloys are one of the leading contenders in this race. In this study, we investigated FA-SMT (or APMT-2), PM-C26M, and Fe17Cr5.5Al over a time period of 6 months in simulated BWR environments and compared their performance with standard Zirc-2 and SS316 materials. Our results implied that water chemistry along with alloy chemistry has a profound effect on the corrosion rate of FeCrAl alloys. Apart from SS316 and Zirc-2 tube specimens, all FeCrAl alloys showed a mass loss in hydrogen water chemistry (HWC). FA-SMT displayed minimal mass loss compared to PM-C26M and Fe17Cr5.5Al because of its higher Cr content. The mass gain of FeCrAl alloys in normal water chemistry (NWC) is significantly less when compared to Zirc-2.
36 MATERIALS SCIENCE↗
Distinctive Oxide Films Develop on the Surface of FeCrAl as the Environment Changes for Nuclear Fuel Cladding
The corrosion-resistant properties of IronChromium–Aluminum (FeCrAl) alloys have been known for nearly a century. Since the 1950s, they have been explored for application in the generation of nuclear power. In the last decade, the focus has been on the use of FeCrAl as cladding for uranium dioxide fuel in light water reactors (LWRs). The corrosion resistance of this alloy depends on the oxide that it can develop on the surface. In LWRs in the vicinity of 300 °C, the external surface oxide of the FeCrAl cladding could be rich in Fe under oxidizing conditions but rich in Cr under reducing conditions. If there is an accident and the cladding is exposed to superheated steam, the cladding will protect itself by developing an alpha aluminum film on the surface.
Qu, Haozheng↗
Influence of grain size on α' Cr precipitation in an isothermally aged Fe-21Cr-5Al alloy
Cr-rich α' precipitation during aging typically leads to hardening and accordingly embrittlement of FeCrAl alloys, which needs to be suppressed. The influence of grain size on α' precipitation was studied by aging coarse-grained (CG), ultra-fine grained (UFG), and nanocrystalline (NC) ferritic Kanthal-D [KD; Fe-21Cr-5Al (wt.%) alloy] at 450, 500 and 550 °C for 500 h. After aging at 450 and 500 °C, less hardening was observed in the UFG KD than in CG KD. Atom probe tomography indicated a lower number density and larger sized intragranular α' in the UFG versus the CG alloy. The smaller grain size and higher defect (vacancy and dislocation) density in the UFG KD facilitated diffusion and accordingly enhanced precipitation kinetics, leading to coarsening of precipitates, as well as saturation of precipitation at lower temperatures, as compared to those in CG KD. No hardening occurred in UFG and CG KD after aging at 550 °C, indicating that the miscibility gap is between 500 and 550 °C. NC KD exhibited softening after aging owing to grain growth. α' precipitation occurred in NC KD aged at 450 °C but not at 500 °C, indicating that miscibility gap is between 450 and 500 °C. Thus, the significantly smaller grain size in NC KD decreased the miscibility gap, as compared to that in CG and UFG KD. Finally, this is attributed to the absorption of vacancies by migrating grain boundaries during aging, suppressing α' nucleation and enhancing Cr solubility.
36 MATERIALS SCIENCE↗
Effect of grain refinement on high temperature steam oxidation of an FeCrAl alloy
FeCrAl alloys are promising candidates to replace Zr alloys as fuel cladding materials in nuclear light-water reactors. Grain refinement has been indicated to improve irradiation resistance. To enhance corrosion resistance as well, the effects of grain refinement on steam corrosion behavior were investigated in this work. Samples of Kanthal D alloy (Fe-21Cr-5Al) with two different grain sizes (coarse-grained and ultrafine-grained) were exposed to steam at 1200 °C for 2 hrs. In conclusion, results indicate improved steam corrosion resistance in ultrafine-grained Kanthal D with formation of a thinner protective Al oxide layer and the presence of a thin underlying Cr oxide layer.
36 MATERIALS SCIENCE↗
Elucidating precipitation in FeCrAl alloys through explainable AI: A case study
Not Available
Materials Science↗
FeCrAl fuel/clad chemical interaction in light water reactor environments
This article investigates the fuel-cladding chemical interaction (FCCI) behavior of two commercial FeCrAl alloys, APMT composition (Fe-21Cr-5Al-3Mo wt.%) and C35M (Fe-13Cr-5Al-2Mo-0.2Si-0.03Y wt.%), after neutron irradiation. “H-cup” diffusion multiples of FeCrAl alloys and ceramic UO 2 fuel were irradiated at a temperature of ~300 °C to a total estimated burnup of 26 GWd/tHM. Post-irradiation Examination results demonstrate the excellent degradation resistance of FeCrAl alloys as accident tolerant fuel (ATF) cladding materials in light water reactor conditions. The study concludes that there was no irradiation-induced defects observed in either of the two commercial FeCrAl claddings. The formation of amorphous Al/U mixed oxide was observed at the fuel-clad interface, which can serve as a tritium permeation barrier and protect against potential chemical attack from the fuel. Finally, the study attributed the formation of amorphous Al/U mixed oxide to the low temperature and limited time of neutron irradiation. APMT forms more distinct Cr and Cr-Fe intermetallic at the FeCrAl-UO 2 interface than C35M due to the higher bulk Cr:Al ratio.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Multi-modal analysis of oxidation on Fe-Cr-Ni austenitic stainless steel 304 exposed to beyond-design basis temperatures
Not Available
Copeland-Johnson, Trishelle M.↗
Optimizing chemistry for designing oxidation resistant FeCrAl alloys
Abstract Traditionally, FeCrAl alloys played an important role in high-temperature applications due to their ability to form a passive Al oxide film at temperatures above ~ 800 °C. Recently, FeCrAl alloys became of interest for the application of accident tolerant nuclear fuel cladding. This study covers work done at GE Research for better understanding the role of Al, Cr, and Mo in oxidation kinetics and thermodynamics. Several models and commercial prototype alloys have been tested in hydrothermal corrosion autoclave loops, at low temperature steam exposure (~ 400 °C), high temperature steam exposure (~ 1000 °C or higher), and high temperature air exposures. The results provide insights on how chromium and aluminum play a significant role in both high temperature and low temperature oxidation of FeCrAl. Additionally, machine learning tools are used to gain further insights on both predicting future optimized chemistries for balancing the properties of hydrothermal corrosion, low and high temperature steam oxidation, and thermal aging (which is exacerbated due to radiation in a nuclear reactor environment). GE plans to use this framework to further optimize the FeCrAl alloy system for use in nuclear reactor environments. Graphical abstract
Roy, Indranil (ORCID:0000000336124323)↗
Experimental investigation of power transient flow boiling
To advance the mechanistic understanding of power transients in flow boiling, experimental investigations are performed to capture the thermal response of four different kinds of cladding materials, including FeCrAls, to Fuchs reactivity initiated accident transients and linear ramp power transients. The transition boiling regime is phenomenologically reflected in power-transient flow boiling, which is different from steady-state boiling. In light of this, the critical heat flux that is usually featured with temperature overshooting does not indicate thermal safety margin of cladding materials as conservatively as the maximum heat flux, which is greater than critical heat flux in the power transient boiling curve. This is physically attributed to the thermal energy deposition on the cladding wall. In addition, the thermal energy released from the cladding wall results in different boiling heat transfer coefficient for the decreasing and increasing power stages respectively during the Fuchs power transient. The maximum heat flux difference gap, which appears due to changes in cladding materials and transient time scales, is appreciable under an intermediate heat convection regime. However, this difference gap is gradually weakened by the progressive increasing of mass flux and/or inlet subcooling due to the enhanced dominance of heat convection over heat conduction. Finally, the small difference gap of maximum heat flux is found to be insignificant under the weak heat convection because of power transient induced annular flow instability.
42 ENGINEERING↗
Effect of surface oxides on tritium entrance and permeation in FeCrAl alloys for nuclear fuel cladding: a review
Abstract Iron-chromium-aluminum (FeCrAl) alloys are being considered for the cladding of uranium dioxide fuel in light water reactors (LWRs). FeCrAl alloys have good mechanical properties at temperatures of 300 °C and higher, and have superlative resistance to attack by steam at temperatures of up to 1000 °C and higher. A concern has been raised that the use of FeCrAl for cladding would result in a higher content of tritium in the reactor coolant as compared with the current system where the cladding is a zirconium based alloy. This review shows that the flux of tritium from the fuel rod cavities to the coolant across the fuel cladding wall will be greatly reduced by the presence of oxides on the surface of the cladding. The review of current literature and permeation data show that (a) protective oxides are expected to be present on both sides of the FeCrAl cladding, and (b) depending on the characteristics of these oxide layers it is reasonable to expect about two–three orders of magnitude reduction in tritium permeation, relative to the permeation response in clean, unoxidized condition for FeCrAl steels of interest, around 277 °C–377 °C temperatures.
Electrochemistry↗
Understanding oxidation of Fe-Cr-Al alloys through explainable artificial intelligence
Abstract The oxidation resistance of FeCrAl based on alloying composition and oxidizing conditions is predicted using a combinatorial experimental and artificial intelligence approach. A neural network (NN) classification model was trained on the experimental FeCrAl dataset produced at GE Research. Furthermore, using the SHapley Additive exPlanations (SHAP) explainable artificial intelligence (XAI) tool, we explore how the NN can showcase further material insights that are unavailable directly from a black-box model. We report that high Al and Cr content forms protective oxide layer, while Mo in FeCrAl creates thick unprotective oxide scale that is vulnerable to spallation due to thermal expansion. Graphical abstract
Materials Science↗