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
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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.",
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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