Surface wettability measurements of FeCrAl alloys under LWR environments

Amir Ali, Maolong Liu, Edward Blandford, Nicholas R. Brown, Kurt A. Terrani, Colby Jensen

    Research output: Contribution to conferencePaperpeer-review

    4 Scopus citations


    Surface wettability analysis, including measurements of contact angle, , and surface roughness, Ra, are effective parameters used in the literature to predict changes in the pool boiling heat transfer Critical Heat Flux (CHF). The CHF is an important aspect of the thermal hydraulic performance that needs to be investigated for new Accident Tolerant Fuel (ATF) cladding materials, such as iron-chromium-aluminum alloys (FeCrAl). This paper presents an investigation of surface wettability of FeCrAl samples oxidized under different simulated Light Water Reactor (LWR) water chemistry conditions. The current investigation includes two major steps: (1) surface roughness and contact angle measurements and CHF prediction based on these measurements which is the main focus of this paper, and (2) conducting atmospheric pool boiling experiments using the same samples to measure CHF. The surface wettability measurements for different samples of as machined (no oxidation) and oxidized FeCrAl were compared to Zircaloy-4 (Zirc-4), the reference cladding material in LWRs and Stainless Steel (310 SS). The obtained results showed no significant difference in the measured contact angle and hence the predicted CHF between the as machined samples of FeCrAl, 310 SS, and Zirc-4. The contact angles for corroded FeCrAl samples under different LWR water chemistry conditions are lower and the measured surface roughness conditions are higher than Zirc-4 corroded under the same conditions. Existing models in the literature predict higher pool boiling CHF of corroded FeCrAl compared to 310 SS, and Zirc-4 under LWR water chemistry autoclave conditions.

    Original languageEnglish (US)
    StatePublished - 2017
    Event17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017 - Xi'an, Shaanxi, China
    Duration: Sep 3 2017Sep 8 2017


    Other17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017
    CityXi'an, Shaanxi

    All Science Journal Classification (ASJC) codes

    • Nuclear Energy and Engineering
    • Instrumentation


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