Critical flow prediction by system codes - Recent analyses made within the FONESYS network

M. Lanfredini (Corresponding author), D. Bestion, F. D'Auria, N. Aksan, P. Fillion, P. Gaillard, J. Heo, I. Karppinen, K. D. Kim, J. Kurki, L. Lifang, A. Shen, J. L. Vacher, D. Wang

    Research output: Chapter in Book/Report/Conference proceedingConference article in proceedingsScientificpeer-review

    Abstract

    A benchmark activity on two-phase critical flow (TPCF) prediction was conducted in the framework of the Forum & Network of System Thermal-Hydraulics Nuclear Reactor Thermal-Hydraulics (FONESYS). FONESYS is a network among code developers who share the common objective to strengthen current technology. The aim of the FONESYS network is to highlight the capabilities and the robustness as well as the limitations of current SYS-TH codes to predict the main phenomena during transient scenarios in nuclear reactors for safety issues. Six separate effect test facilities, more than 100 tests, both in steady and transient conditions, were considered for the activity. Moreover, two ideal tests were designed for code to code comparison in clearly defined conditions. Overall eight system thermal-hydraulic (SYS-TH) codes were adopted, mostly by the developers themselves, ensuring the minimization of the user effect. Results from selected tests were also compared against Delayed Equilibrium Model, not yet implemented in industrial version of SYS-TH codes. Generally, the results of the benchmark show an improvement of the capability of SYS-TH codes to predict TPCF in the last three decades. However, predicting break flowrate remains a major source of uncertainty in accidental transient simulations of LWRs. Progress in understanding of flashing and choked flow might be achieved by physical analysis and setting up meaningful testing programs, but also benefiting from the application of advanced 3-D numerical tools to understand the geometrical effects. Possible ways for improvement of models suitable for SYS-TH codes are discussed.

    Original languageEnglish
    Title of host publication18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 20192019
    PublisherAmerican Nuclear Society ANS
    Pages2274-2287
    Number of pages14
    ISBN (Electronic)978-0-89448-767-5
    Publication statusPublished - 1 Jan 2019
    MoE publication typeA4 Article in a conference publication
    Event18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019 - Portland, United States
    Duration: 18 Aug 201923 Aug 2019

    Conference

    Conference18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019
    Abbreviated titleNURETH-18
    CountryUnited States
    CityPortland
    Period18/08/1923/08/19

    Fingerprint

    critical flow
    hydraulics
    Hydraulics
    predictions
    Nuclear reactors
    photographic developers
    nuclear reactors
    choked flow
    Test facilities
    Hot Temperature
    test facilities
    safety
    Testing
    optimization

    Keywords

    • FONESYS
    • International cooperation
    • SYS-TH codes development
    • Two-phase critical flow

    Cite this

    Lanfredini, M., Bestion, D., D'Auria, F., Aksan, N., Fillion, P., Gaillard, P., ... Wang, D. (2019). Critical flow prediction by system codes - Recent analyses made within the FONESYS network. In 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 20192019 (pp. 2274-2287). American Nuclear Society ANS.
    Lanfredini, M. ; Bestion, D. ; D'Auria, F. ; Aksan, N. ; Fillion, P. ; Gaillard, P. ; Heo, J. ; Karppinen, I. ; Kim, K. D. ; Kurki, J. ; Lifang, L. ; Shen, A. ; Vacher, J. L. ; Wang, D. / Critical flow prediction by system codes - Recent analyses made within the FONESYS network. 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 20192019. American Nuclear Society ANS, 2019. pp. 2274-2287
    @inproceedings{411d9137372147dab3b7278bd2c1d98b,
    title = "Critical flow prediction by system codes - Recent analyses made within the FONESYS network",
    abstract = "A benchmark activity on two-phase critical flow (TPCF) prediction was conducted in the framework of the Forum & Network of System Thermal-Hydraulics Nuclear Reactor Thermal-Hydraulics (FONESYS). FONESYS is a network among code developers who share the common objective to strengthen current technology. The aim of the FONESYS network is to highlight the capabilities and the robustness as well as the limitations of current SYS-TH codes to predict the main phenomena during transient scenarios in nuclear reactors for safety issues. Six separate effect test facilities, more than 100 tests, both in steady and transient conditions, were considered for the activity. Moreover, two ideal tests were designed for code to code comparison in clearly defined conditions. Overall eight system thermal-hydraulic (SYS-TH) codes were adopted, mostly by the developers themselves, ensuring the minimization of the user effect. Results from selected tests were also compared against Delayed Equilibrium Model, not yet implemented in industrial version of SYS-TH codes. Generally, the results of the benchmark show an improvement of the capability of SYS-TH codes to predict TPCF in the last three decades. However, predicting break flowrate remains a major source of uncertainty in accidental transient simulations of LWRs. Progress in understanding of flashing and choked flow might be achieved by physical analysis and setting up meaningful testing programs, but also benefiting from the application of advanced 3-D numerical tools to understand the geometrical effects. Possible ways for improvement of models suitable for SYS-TH codes are discussed.",
    keywords = "FONESYS, International cooperation, SYS-TH codes development, Two-phase critical flow",
    author = "M. Lanfredini and D. Bestion and F. D'Auria and N. Aksan and P. Fillion and P. Gaillard and J. Heo and I. Karppinen and Kim, {K. D.} and J. Kurki and L. Lifang and A. Shen and Vacher, {J. L.} and D. Wang",
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    language = "English",
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    Lanfredini, M, Bestion, D, D'Auria, F, Aksan, N, Fillion, P, Gaillard, P, Heo, J, Karppinen, I, Kim, KD, Kurki, J, Lifang, L, Shen, A, Vacher, JL & Wang, D 2019, Critical flow prediction by system codes - Recent analyses made within the FONESYS network. in 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 20192019. American Nuclear Society ANS, pp. 2274-2287, 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, Portland, United States, 18/08/19.

    Critical flow prediction by system codes - Recent analyses made within the FONESYS network. / Lanfredini, M. (Corresponding author); Bestion, D.; D'Auria, F.; Aksan, N.; Fillion, P.; Gaillard, P.; Heo, J.; Karppinen, I.; Kim, K. D.; Kurki, J.; Lifang, L.; Shen, A.; Vacher, J. L.; Wang, D.

    18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 20192019. American Nuclear Society ANS, 2019. p. 2274-2287.

    Research output: Chapter in Book/Report/Conference proceedingConference article in proceedingsScientificpeer-review

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    T1 - Critical flow prediction by system codes - Recent analyses made within the FONESYS network

    AU - Lanfredini, M.

    AU - Bestion, D.

    AU - D'Auria, F.

    AU - Aksan, N.

    AU - Fillion, P.

    AU - Gaillard, P.

    AU - Heo, J.

    AU - Karppinen, I.

    AU - Kim, K. D.

    AU - Kurki, J.

    AU - Lifang, L.

    AU - Shen, A.

    AU - Vacher, J. L.

    AU - Wang, D.

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    N2 - A benchmark activity on two-phase critical flow (TPCF) prediction was conducted in the framework of the Forum & Network of System Thermal-Hydraulics Nuclear Reactor Thermal-Hydraulics (FONESYS). FONESYS is a network among code developers who share the common objective to strengthen current technology. The aim of the FONESYS network is to highlight the capabilities and the robustness as well as the limitations of current SYS-TH codes to predict the main phenomena during transient scenarios in nuclear reactors for safety issues. Six separate effect test facilities, more than 100 tests, both in steady and transient conditions, were considered for the activity. Moreover, two ideal tests were designed for code to code comparison in clearly defined conditions. Overall eight system thermal-hydraulic (SYS-TH) codes were adopted, mostly by the developers themselves, ensuring the minimization of the user effect. Results from selected tests were also compared against Delayed Equilibrium Model, not yet implemented in industrial version of SYS-TH codes. Generally, the results of the benchmark show an improvement of the capability of SYS-TH codes to predict TPCF in the last three decades. However, predicting break flowrate remains a major source of uncertainty in accidental transient simulations of LWRs. Progress in understanding of flashing and choked flow might be achieved by physical analysis and setting up meaningful testing programs, but also benefiting from the application of advanced 3-D numerical tools to understand the geometrical effects. Possible ways for improvement of models suitable for SYS-TH codes are discussed.

    AB - A benchmark activity on two-phase critical flow (TPCF) prediction was conducted in the framework of the Forum & Network of System Thermal-Hydraulics Nuclear Reactor Thermal-Hydraulics (FONESYS). FONESYS is a network among code developers who share the common objective to strengthen current technology. The aim of the FONESYS network is to highlight the capabilities and the robustness as well as the limitations of current SYS-TH codes to predict the main phenomena during transient scenarios in nuclear reactors for safety issues. Six separate effect test facilities, more than 100 tests, both in steady and transient conditions, were considered for the activity. Moreover, two ideal tests were designed for code to code comparison in clearly defined conditions. Overall eight system thermal-hydraulic (SYS-TH) codes were adopted, mostly by the developers themselves, ensuring the minimization of the user effect. Results from selected tests were also compared against Delayed Equilibrium Model, not yet implemented in industrial version of SYS-TH codes. Generally, the results of the benchmark show an improvement of the capability of SYS-TH codes to predict TPCF in the last three decades. However, predicting break flowrate remains a major source of uncertainty in accidental transient simulations of LWRs. Progress in understanding of flashing and choked flow might be achieved by physical analysis and setting up meaningful testing programs, but also benefiting from the application of advanced 3-D numerical tools to understand the geometrical effects. Possible ways for improvement of models suitable for SYS-TH codes are discussed.

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    Lanfredini M, Bestion D, D'Auria F, Aksan N, Fillion P, Gaillard P et al. Critical flow prediction by system codes - Recent analyses made within the FONESYS network. In 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 20192019. American Nuclear Society ANS. 2019. p. 2274-2287