Surface wettability and pool boiling Critical Heat Flux of Accident Tolerant Fuel cladding-FeCrAl alloys

Amir F. Ali, Jacob P. Gorton, Nicholas Brown, Kurt A. Terrani, Colby B. Jensen, Youho Lee, Edward D. Blandford

    Research output: Contribution to journalArticle

    6 Citations (Scopus)

    Abstract

    Surface wettability analysis, including measurements of static (θ) advance (θA), and receding (θr) contact angles, and surface roughness, Ra, are effective parameters used in the literature to predict changes in the pool boiling 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). Surface wettability of FeCrAl samples oxidized under different simulated Light Water Reactor (LWR) water chemistry conditions was measured. These measurements were compared to as-machined and oxidized Zircaloy-4 (Zirc-4), the reference cladding material in LWRs, under the same conditions. Theoretical models were used to predict the pool boiling CHF using surface wettability measurements. Pool boiling experiments were conducted using the same samples to measure the CHF. The measured and predicted CHF data were compared for model validation. The obtained results showed no significant difference in the measured static contact angle and hence the predicted CHF between the as-machined samples of FeCrAl, 310 SS, and Zirc-4. The contact angles (static, advance, and receding angles) for corroded FeCrAl samples under different LWR water chemistry conditions are lower, and the measured surface roughness values are higher than Zirc-4 corroded under the same conditions. Existing models in the literature predicted higher pool boiling CHF of corroded FeCrAl compared to 310 stainless steel (SS), and Zirc-4. The measured pool boiling CHF for oxidized FeCrAl samples was higher than Zirc-4 samples. The measured and predicted CHF values are in good agreement. The CHF and the Departure from Nucleate Boiling Ratio (DNBR) were calculated using the Consortium for Advanced Simulation of Light Water Reactors (CASL) subchannel code COBRA-TF (CTF) for a one-eighth model of a 17 × 17 Pressurized Water Reactor (PWR) fuel assembly. The inlet conditions are consistent with typical PWR values except for the power, which was set 50% higher than is typical to represent possible accident conditions more accurately. The calculated distributions for the CHF and DNBR for oxidized FeCrAl showed significantly higher values throughout the fuel assembly octant compared to as machined Zirc-4 and as machined FeCrAl. These preliminary results show that oxidized FeCrAl may be able to withstand the proposed accident conditions without leading to a boiling crisis.

    Original languageEnglish (US)
    Pages (from-to)218-231
    Number of pages14
    JournalNuclear Engineering and Design
    Volume338
    DOIs
    StatePublished - Nov 1 2018

    Fingerprint

    wettability
    accidents
    boiling
    Boiling liquids
    heat flux
    accident
    Wetting
    Heat flux
    Accidents
    light water reactors
    Light water reactors
    Contact angle
    pressurized water reactors
    nucleate boiling
    Nucleate boiling
    Pressurized water reactors
    Stainless Steel
    surface roughness
    water chemistry
    stainless steels

    All Science Journal Classification (ASJC) codes

    • Nuclear and High Energy Physics
    • Nuclear Energy and Engineering
    • Materials Science(all)
    • Safety, Risk, Reliability and Quality
    • Waste Management and Disposal
    • Mechanical Engineering

    Cite this

    Ali, A. F., Gorton, J. P., Brown, N., Terrani, K. A., Jensen, C. B., Lee, Y., & Blandford, E. D. (2018). Surface wettability and pool boiling Critical Heat Flux of Accident Tolerant Fuel cladding-FeCrAl alloys. Nuclear Engineering and Design, 338, 218-231. https://doi.org/10.1016/j.nucengdes.2018.08.024
    Ali, Amir F. ; Gorton, Jacob P. ; Brown, Nicholas ; Terrani, Kurt A. ; Jensen, Colby B. ; Lee, Youho ; Blandford, Edward D. / Surface wettability and pool boiling Critical Heat Flux of Accident Tolerant Fuel cladding-FeCrAl alloys. In: Nuclear Engineering and Design. 2018 ; Vol. 338. pp. 218-231.
    @article{3eb6896767ea424aaa0590b011146640,
    title = "Surface wettability and pool boiling Critical Heat Flux of Accident Tolerant Fuel cladding-FeCrAl alloys",
    abstract = "Surface wettability analysis, including measurements of static (θ) advance (θA), and receding (θr) contact angles, and surface roughness, Ra, are effective parameters used in the literature to predict changes in the pool boiling 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). Surface wettability of FeCrAl samples oxidized under different simulated Light Water Reactor (LWR) water chemistry conditions was measured. These measurements were compared to as-machined and oxidized Zircaloy-4 (Zirc-4), the reference cladding material in LWRs, under the same conditions. Theoretical models were used to predict the pool boiling CHF using surface wettability measurements. Pool boiling experiments were conducted using the same samples to measure the CHF. The measured and predicted CHF data were compared for model validation. The obtained results showed no significant difference in the measured static contact angle and hence the predicted CHF between the as-machined samples of FeCrAl, 310 SS, and Zirc-4. The contact angles (static, advance, and receding angles) for corroded FeCrAl samples under different LWR water chemistry conditions are lower, and the measured surface roughness values are higher than Zirc-4 corroded under the same conditions. Existing models in the literature predicted higher pool boiling CHF of corroded FeCrAl compared to 310 stainless steel (SS), and Zirc-4. The measured pool boiling CHF for oxidized FeCrAl samples was higher than Zirc-4 samples. The measured and predicted CHF values are in good agreement. The CHF and the Departure from Nucleate Boiling Ratio (DNBR) were calculated using the Consortium for Advanced Simulation of Light Water Reactors (CASL) subchannel code COBRA-TF (CTF) for a one-eighth model of a 17 × 17 Pressurized Water Reactor (PWR) fuel assembly. The inlet conditions are consistent with typical PWR values except for the power, which was set 50{\%} higher than is typical to represent possible accident conditions more accurately. The calculated distributions for the CHF and DNBR for oxidized FeCrAl showed significantly higher values throughout the fuel assembly octant compared to as machined Zirc-4 and as machined FeCrAl. These preliminary results show that oxidized FeCrAl may be able to withstand the proposed accident conditions without leading to a boiling crisis.",
    author = "Ali, {Amir F.} and Gorton, {Jacob P.} and Nicholas Brown and Terrani, {Kurt A.} and Jensen, {Colby B.} and Youho Lee and Blandford, {Edward D.}",
    year = "2018",
    month = "11",
    day = "1",
    doi = "10.1016/j.nucengdes.2018.08.024",
    language = "English (US)",
    volume = "338",
    pages = "218--231",
    journal = "Nuclear Engineering and Design",
    issn = "0029-5493",
    publisher = "Elsevier BV",

    }

    Surface wettability and pool boiling Critical Heat Flux of Accident Tolerant Fuel cladding-FeCrAl alloys. / Ali, Amir F.; Gorton, Jacob P.; Brown, Nicholas; Terrani, Kurt A.; Jensen, Colby B.; Lee, Youho; Blandford, Edward D.

    In: Nuclear Engineering and Design, Vol. 338, 01.11.2018, p. 218-231.

    Research output: Contribution to journalArticle

    TY - JOUR

    T1 - Surface wettability and pool boiling Critical Heat Flux of Accident Tolerant Fuel cladding-FeCrAl alloys

    AU - Ali, Amir F.

    AU - Gorton, Jacob P.

    AU - Brown, Nicholas

    AU - Terrani, Kurt A.

    AU - Jensen, Colby B.

    AU - Lee, Youho

    AU - Blandford, Edward D.

    PY - 2018/11/1

    Y1 - 2018/11/1

    N2 - Surface wettability analysis, including measurements of static (θ) advance (θA), and receding (θr) contact angles, and surface roughness, Ra, are effective parameters used in the literature to predict changes in the pool boiling 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). Surface wettability of FeCrAl samples oxidized under different simulated Light Water Reactor (LWR) water chemistry conditions was measured. These measurements were compared to as-machined and oxidized Zircaloy-4 (Zirc-4), the reference cladding material in LWRs, under the same conditions. Theoretical models were used to predict the pool boiling CHF using surface wettability measurements. Pool boiling experiments were conducted using the same samples to measure the CHF. The measured and predicted CHF data were compared for model validation. The obtained results showed no significant difference in the measured static contact angle and hence the predicted CHF between the as-machined samples of FeCrAl, 310 SS, and Zirc-4. The contact angles (static, advance, and receding angles) for corroded FeCrAl samples under different LWR water chemistry conditions are lower, and the measured surface roughness values are higher than Zirc-4 corroded under the same conditions. Existing models in the literature predicted higher pool boiling CHF of corroded FeCrAl compared to 310 stainless steel (SS), and Zirc-4. The measured pool boiling CHF for oxidized FeCrAl samples was higher than Zirc-4 samples. The measured and predicted CHF values are in good agreement. The CHF and the Departure from Nucleate Boiling Ratio (DNBR) were calculated using the Consortium for Advanced Simulation of Light Water Reactors (CASL) subchannel code COBRA-TF (CTF) for a one-eighth model of a 17 × 17 Pressurized Water Reactor (PWR) fuel assembly. The inlet conditions are consistent with typical PWR values except for the power, which was set 50% higher than is typical to represent possible accident conditions more accurately. The calculated distributions for the CHF and DNBR for oxidized FeCrAl showed significantly higher values throughout the fuel assembly octant compared to as machined Zirc-4 and as machined FeCrAl. These preliminary results show that oxidized FeCrAl may be able to withstand the proposed accident conditions without leading to a boiling crisis.

    AB - Surface wettability analysis, including measurements of static (θ) advance (θA), and receding (θr) contact angles, and surface roughness, Ra, are effective parameters used in the literature to predict changes in the pool boiling 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). Surface wettability of FeCrAl samples oxidized under different simulated Light Water Reactor (LWR) water chemistry conditions was measured. These measurements were compared to as-machined and oxidized Zircaloy-4 (Zirc-4), the reference cladding material in LWRs, under the same conditions. Theoretical models were used to predict the pool boiling CHF using surface wettability measurements. Pool boiling experiments were conducted using the same samples to measure the CHF. The measured and predicted CHF data were compared for model validation. The obtained results showed no significant difference in the measured static contact angle and hence the predicted CHF between the as-machined samples of FeCrAl, 310 SS, and Zirc-4. The contact angles (static, advance, and receding angles) for corroded FeCrAl samples under different LWR water chemistry conditions are lower, and the measured surface roughness values are higher than Zirc-4 corroded under the same conditions. Existing models in the literature predicted higher pool boiling CHF of corroded FeCrAl compared to 310 stainless steel (SS), and Zirc-4. The measured pool boiling CHF for oxidized FeCrAl samples was higher than Zirc-4 samples. The measured and predicted CHF values are in good agreement. The CHF and the Departure from Nucleate Boiling Ratio (DNBR) were calculated using the Consortium for Advanced Simulation of Light Water Reactors (CASL) subchannel code COBRA-TF (CTF) for a one-eighth model of a 17 × 17 Pressurized Water Reactor (PWR) fuel assembly. The inlet conditions are consistent with typical PWR values except for the power, which was set 50% higher than is typical to represent possible accident conditions more accurately. The calculated distributions for the CHF and DNBR for oxidized FeCrAl showed significantly higher values throughout the fuel assembly octant compared to as machined Zirc-4 and as machined FeCrAl. These preliminary results show that oxidized FeCrAl may be able to withstand the proposed accident conditions without leading to a boiling crisis.

    UR - http://www.scopus.com/inward/record.url?scp=85052324027&partnerID=8YFLogxK

    UR - http://www.scopus.com/inward/citedby.url?scp=85052324027&partnerID=8YFLogxK

    U2 - 10.1016/j.nucengdes.2018.08.024

    DO - 10.1016/j.nucengdes.2018.08.024

    M3 - Article

    AN - SCOPUS:85052324027

    VL - 338

    SP - 218

    EP - 231

    JO - Nuclear Engineering and Design

    JF - Nuclear Engineering and Design

    SN - 0029-5493

    ER -