Nuclear Power Engineering, Volume. 46, Issue 4, 1(2025)
Development Progress of CorTAF: A Three-Dimensional Cross-Scale Multi-Physics Coupling Analysis Code for Nuclear Reactor Core
[1] [1] KULESZA J A, FRANCESCHINI F, EVANS T M, et al. Overview of the consortium for the advanced simulation of light water reactors (CASL)[J]. EPJ Web of Conferences, 2016, 106: 03002.
[2] [2] SZILARD R, KOTHE D, TURINSKY P. The consortium for advanced simulation of light water reactors[C]//Enlarged Halden Programme Group Meeting. Sandefjord, Norway: INL, 2011.
[3] [3] LEFEBVRE R A, LANGLEY B R, MILLER L P, et al. NEAMS workbench status and capabilities: ORNL/TM-2019/1314[R]. Tennessee: Oak Ridge National Laboratory (ORNL), 2019.
[4] [4] CHANARON B. Overview of the NURESAFE European project[J]. Nuclear Engineering and Design, 2017, 321: 1-7.
[5] [5] CHANARON B, AHNERT C, CROUZET N, et al. Advanced multi-physics simulation for reactor safety in the framework of the NURESAFE project[J]. Annals of Nuclear Energy, 2015, 84: 166-177.
[6] [6] CHAULIAC C, ARAGONS J M, BESTION D, et al. NURESIM – a European simulation platform for nuclear reactor safety: multi-scale and multi-physics calculations, sensitivity and uncertainty analysis[J]. Nuclear Engineering and Design, 2011, 241(9): 3416-3426.
[11] [11] DONG Z Y, LIU K, WANG M J, et al. Development of Two-Dimensional heat conduction model in subchannel code based on OpenFOAM[C].//Proceedings of the 2024 31st International Conference on Nuclear Engineering. Prague, Czech Republic: Nuclear Engineering Division, 2024.
[12] [12] DONG Z Y, LIU K, QIU H R, et al. Preliminary implementation of high-resolution multi-scale coupling calculations for the entire pressure vessel based on OpenFOAM[J]. Applied Thermal Engineering, 2025, 259: 124911.
[13] [13] LIU K, WANG M J, TIAN W X, et al. CorTAF: a nuclear reactor core three-dimensional thermal-hydraulic characteristics analysis code based on OpenFOAM[J]. Nuclear Engineering and Design, 2023, 405: 112209.
[14] [14] LIU X T, LIU K, WANG M J, et al. Three-Dimensional thermal-hydraulic characteristics analysis of plate-type fuel reactor core based on OpenFOAM[J]. Progress in Nuclear Energy, 2023, 160: 104712.
[15] [15] QIU H R, YU J C, WANG M J, et al. Application of BPNN algorithm in thermal-hydraulic analysis of unwrapped LFR core[J]. International Journal of Thermal Sciences, 2024, 203: 109176.
[16] [16] YU J C, LIU K, QIU H R, et al. Thermal-hydraulic analysis of a full-scale lead-bismuth cooled fast reactor core considering inter-wrapper flow[J]. Progress in Nuclear Energy, 2024, 175: 105331.
[17] [17] YU J C, LIU K, QIU H R, et al. Numerical analysis of thermal-hydraulic characteristics of the whole LFR core under blockage conditions[J]. International Journal of Thermal Sciences, 2025, 208: 109435.
[18] [18] DONG Z Y, LIU K, WANG M J, et al. The development of nuclear reactor three-dimensional neutronic thermal–hydraulic coupling code: CorTAF-2.0[J]. Nuclear Engineering and Design, 2023, 415: 112689.
[19] [19] DONG Z Y, LIU K, WANG M J, et al. Study on the deposition migration and heat transfer characteristics in the reactor core based on OpenFOAM[J]. Applied Thermal Engineering, 2023, 230: 120858.
[20] [20] GE J, ZHANG D L, TIAN W X, et al. Steady and transient solutions of neutronics problems based on finite volume method (FVM) with a CFD code[J]. Progress in Nuclear Energy, 2015, 85: 366-374.
[21] [21] HUO Y C, YU H, WANG M J, et al. Development and application of TaSNAM 2.0 for advanced pressurized water reactor[J]. Annals of Nuclear Energy, 2022, 166: 108801.
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Su Guanghui, Dong Zhengyang, Liu Kai, Wang Mingjun, Tian Wenxi, Qiu Suizheng. Development Progress of CorTAF: A Three-Dimensional Cross-Scale Multi-Physics Coupling Analysis Code for Nuclear Reactor Core[J]. Nuclear Power Engineering, 2025, 46(4): 1
Received: May. 9, 2025
Accepted: Aug. 25, 2025
Published Online: Aug. 25, 2025
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