Experimental and Numerical Characterization of the Flow Field at the Core Entrance of a Water Model of a Heavy Liquid Metal–Cooled Reactor
Research output: Contribution to journal › Article › peer-review
Standard
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.Research output: Contribution to journal › Article › peer-review
Harvard
APA
Vancouver
Author
Bibtex - Download
}
RIS - Download
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