Experimental and Numerical Characterization of the Flow Field at the Core Entrance of a Water Model of a Heavy Liquid Metal–Cooled Reactor

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Experimental and Numerical Characterization of the Flow Field at the Core Entrance of a Water Model of a Heavy Liquid Metal–Cooled Reactor. / Planquart, Philippe; Spaccapaniccia, Chiara; Alessi, Giacomo; Buckingham, Sophia; Van Tichelen, Katrien.

In: Nuclear Technology, Vol. 206, No. 2, 02.2020, p. 231-241.

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Planquart, Philippe ; Spaccapaniccia, Chiara ; Alessi, Giacomo ; Buckingham, Sophia ; Van Tichelen, Katrien. / Experimental and Numerical Characterization of the Flow Field at the Core Entrance of a Water Model of a Heavy Liquid Metal–Cooled Reactor. In: Nuclear Technology. 2020 ; Vol. 206, No. 2. pp. 231-241.

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@article{8a29e329b9cf4928bbbc6acaad997469,
title = "Experimental and Numerical Characterization of the Flow Field at the Core Entrance of a Water Model of a Heavy Liquid Metal–Cooled Reactor",
abstract = "The thermal-hydraulic challenges of a nuclear reactor are numerous and mastering them is crucial for the design and safety of new reactors. Numerical simulation through computational fluid dynamics (CFD) codes or system thermal-hydraulic codes can address a lot of the different questions, nevertheless the use of water modeling for the study of the thermal-hydraulic behavior of a new primary system and the validation of codes remains an extremely valuable tool. A water model of the pool-type PbBi-cooled MYRRHA reactor has been developed at the von Karman Institute in collaboration with SCK•CEN. It is a full plexiglass model at a geometrical scale 1/5 of MYRRHA. This transparent water model allows the application of optical measurement techniques like particle image velocimetry (PIV) for flow characterization. Local results of PIV measurements performed in the lower plenum at the entrance of the core are presented and compared with CFD results for nominal operating condition and a natural convection case simulating decay heat removal. Very good agreement has been found in the velocity field. The results also show the importance of the radial flow entering the core of the water model in natural convection.",
keywords = "MYRRHA, water model, particle image velocimetry, computational fluid dynamics, validation",
author = "Philippe Planquart and Chiara Spaccapaniccia and Giacomo Alessi and Sophia Buckingham and {Van Tichelen}, Katrien",
note = "Score=10; 2018 - ATH - Advances in Thermal Hydraulics meeting ; Conference date: 11-11-2018 Through 15-11-2018",
year = "2020",
month = feb,
doi = "10.1080/00295450.2019.1637240",
language = "English",
volume = "206",
pages = "231--241",
journal = "Nuclear Technology",
issn = "0029-5450",
publisher = "Springer",
number = "2",

}

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TY - JOUR

T1 - Experimental and Numerical Characterization of the Flow Field at the Core Entrance of a Water Model of a Heavy Liquid Metal–Cooled Reactor

AU - Planquart, Philippe

AU - Spaccapaniccia, Chiara

AU - Alessi, Giacomo

AU - Buckingham, Sophia

AU - Van Tichelen, Katrien

N1 - Score=10

PY - 2020/2

Y1 - 2020/2

N2 - The thermal-hydraulic challenges of a nuclear reactor are numerous and mastering them is crucial for the design and safety of new reactors. Numerical simulation through computational fluid dynamics (CFD) codes or system thermal-hydraulic codes can address a lot of the different questions, nevertheless the use of water modeling for the study of the thermal-hydraulic behavior of a new primary system and the validation of codes remains an extremely valuable tool. A water model of the pool-type PbBi-cooled MYRRHA reactor has been developed at the von Karman Institute in collaboration with SCK•CEN. It is a full plexiglass model at a geometrical scale 1/5 of MYRRHA. This transparent water model allows the application of optical measurement techniques like particle image velocimetry (PIV) for flow characterization. Local results of PIV measurements performed in the lower plenum at the entrance of the core are presented and compared with CFD results for nominal operating condition and a natural convection case simulating decay heat removal. Very good agreement has been found in the velocity field. The results also show the importance of the radial flow entering the core of the water model in natural convection.

AB - The thermal-hydraulic challenges of a nuclear reactor are numerous and mastering them is crucial for the design and safety of new reactors. Numerical simulation through computational fluid dynamics (CFD) codes or system thermal-hydraulic codes can address a lot of the different questions, nevertheless the use of water modeling for the study of the thermal-hydraulic behavior of a new primary system and the validation of codes remains an extremely valuable tool. A water model of the pool-type PbBi-cooled MYRRHA reactor has been developed at the von Karman Institute in collaboration with SCK•CEN. It is a full plexiglass model at a geometrical scale 1/5 of MYRRHA. This transparent water model allows the application of optical measurement techniques like particle image velocimetry (PIV) for flow characterization. Local results of PIV measurements performed in the lower plenum at the entrance of the core are presented and compared with CFD results for nominal operating condition and a natural convection case simulating decay heat removal. Very good agreement has been found in the velocity field. The results also show the importance of the radial flow entering the core of the water model in natural convection.

KW - MYRRHA

KW - water model

KW - particle image velocimetry

KW - computational fluid dynamics

KW - validation

UR - http://ecm.sckcen.be/OTCS/llisapi.dll/open/36948188

U2 - 10.1080/00295450.2019.1637240

DO - 10.1080/00295450.2019.1637240

M3 - Article

VL - 206

SP - 231

EP - 241

JO - Nuclear Technology

JF - Nuclear Technology

SN - 0029-5450

IS - 2

T2 - 2018 - ATH - Advances in Thermal Hydraulics meeting

Y2 - 11 November 2018 through 15 November 2018

ER -

ID: 5973940