Predicting Critical Heat Flux in Narrow Rectangular Channels: A CFD Approach for Subcooled Flow

被引:0
|
作者
Bertocchi, F. [1 ]
Vega, C. A. [2 ]
机构
[1] NRG PALLAS, Utrechtseweg 310, NL-6812 AR Arnhem, Netherlands
[2] NRG PALLAS, Comeniusstr 8, NL-1817 MS Alkmaar, Netherlands
关键词
Critical heat flux; Rectangular channel; Computational fluid dynamics; Deterministic safety analyses; PALLAS-reactor; CHF CORRELATION; MODELS;
D O I
10.1016/j.nucengdes.2024.113740
中图分类号
TL [原子能技术]; O571 [原子核物理学];
学科分类号
0827 ; 082701 ;
摘要
The PALLAS-reactor is an advanced nuclear reactor designed for producing medical isotopes and for carrying out research on (medical) nuclear technology. RELAP5 is the primary tool for modeling the accident scenarios relevant to the licensing of this reactor. The Groeneveld Look-Up Tables (LUT) are the default model of RELAP5 for estimating Critical Heat Flux (CHF). Literature has shown that the LUT over-predict CHF inside narrow, rectangular channels such as the cooling channels of the core of the PALLAS-reactor. Empirical correlations such as that of Sudo-Kaminaga or its modified version by Kim et al. (SK-Kim) are better suited for these geometries. However, these correlations are typically implemented in one-dimensional (1D) System Thermal-Hydraulics codes whereby the quantities of interest are averaged-out over the control domains. Computational Fluid Dynamics (CFD) could capture three-dimensional effects that could help reducing unnecessary conservatism embedded in some 1D approaches (e.g. Sudo-Kaminaga or SK-Kim correlations). Therefore, CFD is to be employed to support the Deterministic Safety Analyses of the PALLAS-reactor if there are scenarios in which the SK-Kim correlation predicts CHF. Boiling flows inside round or annular vertical channels have been extensively modeled with CFD. However, attempts at modeling CHF in narrow, rectangular geometries are still scarce. Therefore, this work aims at developing and validating a CFD approach for predicting CHF in rectangular channels. Four combinations of turbulence interaction and turbulent dispersion models are investigated; the results are compared against experiments and the SK-Kim correlation; sensitivity analyses are conducted on the turbulence models as well as the mesh. Certain combinations of turbulence interaction and turbulent dispersion models promote the transport of steam towards the center of the channel, thus increasing CHF, while other cause the steam to remain confined at the wall, thereby lowering the CHF. This study provides several CFD options to estimate CHF depending on the required level of conservatism. Estimating CHF in a conservative manner is crucial for licensing the PALLAS-reactor and other reactors of similar design, thereby securing the production of medical isotopes in Europe and the rest of the world.
引用
收藏
页数:15
相关论文
共 50 条
  • [31] Critical heat flux for subcooled flow in an annulus with bilateral heating
    Chen, YZ
    Zhou, RB
    Chen, HY
    HEAT TRANSFER SCIENCE AND TECHNOLOGY 1996, 1996, : 376 - 381
  • [32] Viewpoint of subcooled flow boiling critical heat flux mechanism
    Liu, W
    Nariai, H
    CHEMICAL ENGINEERING & TECHNOLOGY, 2002, 25 (04) : 447 - 453
  • [33] Analysis of the critical heat flux of subcooled flow boiling in microgravity
    Liu, Bin
    Yuan, Bo
    Zhou, Jie
    Zhao, Jianfu
    Di Marco, Paolo
    Zhang, Yonghai
    Wei, Jinjia
    Yang, Yang
    EXPERIMENTAL THERMAL AND FLUID SCIENCE, 2021, 120
  • [34] Viewpoint of subcooled flow boiling critical heat flux mechanism
    Liu, W.
    Nariai, H.
    Chemical Engineering and Technology, 2002, 25 (04): : 447 - 453
  • [35] A critical heat flux experiment with water flow at low pressures in thin rectangular channels
    Lu, Donghua
    Wen, Qinglong
    Liu, Tong
    Su, Qianhua
    NUCLEAR ENGINEERING AND DESIGN, 2014, 278 : 669 - 678
  • [36] Prediction of critical heat flux for narrow rectangular channels in a steady state condition using machine learning
    Kim, Huiyung
    Moon, Jeongmin
    Hong, Dongjin
    Cha, Euiyoung
    Yun, Byongjo
    NUCLEAR ENGINEERING AND TECHNOLOGY, 2021, 53 (06) : 1796 - 1809
  • [37] A study on prediction methods of the critical heat flux for upward flow in a vertical narrow rectangular channel
    Choi, Gil Sik
    Jeong, Yong Hoon
    Chang, Soon Heung
    NUCLEAR ENGINEERING AND DESIGN, 2015, 294 : 103 - 116
  • [38] An evaluation of critical heat flux prediction methods for the upward flow in a vertical narrow rectangular channel
    Yan, Meiyue
    Ma, Zaiyong
    Pan, Liangming
    Liu, Wei
    He, Qingche
    Zhang, Rui
    Wu, Qi
    Xu, Wangtao
    PROGRESS IN NUCLEAR ENERGY, 2021, 140
  • [39] Numerical investigation of critical heat flux during subcooled flow boiling in a vertical rectangular Mini-channel
    Chen, Yu-Jie
    Ling, Kong
    Jin, Shu-Qi
    Lu, Wei
    Yu, Bo
    Sun, Dongliang
    Zhang, Wei
    Tao, Wen-Quan
    APPLIED THERMAL ENGINEERING, 2023, 221
  • [40] Prediction of the critical heat flux in water subcooled flow boiling using a new mechanistic approach
    Celata, GP
    Cumo, M
    Katto, Y
    Mariani, A
    INTERNATIONAL JOURNAL OF HEAT AND MASS TRANSFER, 1999, 42 (08) : 1457 - 1466