[1] ZHU Jizhou, XI Shuren, SHAN Jianqiang et al[M]. Nuclear reactor safety analysis(2000).

[2] Markus E, Dietrich K, Antonio H et al. Numerical discretization analysis of a HTR steam generator model for the thermal-hydraulics code trace[J]. Nuclear Technology and Radiation Protection, 29, 31-38(2014).

[3] Yoon J, Kim J P, Kim H Y et al. Development of a computer code, ONCESG, for the thermal-hydraulic design of a once-through steam generator[J]. Journal of Nuclear Science and Technology, 37, 445-454(2000).

[4] Hwang Y D, Yang S H, Kim S H et al. Model description of TASS/SMR code, KATRI/TR-3082/2005[R](2005).

[5] Ma Y, Li X W, Wu X X. Thermal–hydraulic characteristics and flow instability analysis of an HTGR helical tube steam generator[J]. Annals of Nuclear Energy, 73, 484-495(2014).

[6] Wu Y J, Liu B K, Zhang H et al. A movable boundary model for helical coiled once-through steam generator using preconditioned JFNK method[J]. International Journal of Advanced Nuclear Reactor Design and Technology, 4, 1-8(2022).

[7] YUAN Yuan. Modeling and simulation of spiral tube direct flow steam generator[D](2015).

[8] WANG Mingjun, SUN Jinxiang, ZHANG Jing et al. Thermal-hydraulic characteristics analysis method of helical coil steam generator in high temperature gas reactor[J]. Atomic Energy Science and Technology, 56, 2262-2271(2022).

[9] WU Pan, OUYANG Zeyu, ZHU Yu et al. Validation of an inhouse system analysis code for heat pipe cooled reactor[J]. Nuclear Techniques, 46, 110603(2023).

[10] LIU Dan, SUN Jun, SUN Yuliang. HTR-PM once-through steam generator modeling and analysis[J]. Atomic Energy Science and Technology, 50, 7(2016).

[11] Itō H. Friction factors for turbulent flow in curved pipes[J]. Journal of Basic Engineering, 81, 123-132(1959).

[12] Cioncolini A, Santini L. Two-phase pressure drop prediction in helically coiled steam generators for nuclear power applications[J]. International Journal of Heat and Mass Transfer, 100, 825-834(2016).

[13] Gilli P V. Heat transfer and pressure drop for cross flow through banks of multistart helical tubes with uniform inclinations and uniform longitudinal pitches[J]. Nuclear Science and Engineering, 22, 298-314(1965).

[14] Schmidt E F. Wärmeübergang und druckverlust in rohrschlangen[J]. Chemie Ingenieur Technik, 39, 781-899(1967).

[15] Rogers G F C, Mayhew Y R. Heat transfer and pressure loss in helically coiled tubes with turbulent flow[J]. International Journal of Heat and Mass Transfer, 7, 1207-1216(1964).

[16] XU Jiyun[M]. Boiling heat transfer and gas-liquid two-phase flow, 488-491(2000).

[17] Miropolskiy Z L. Heat transfer in boiling of a steam-water mixture in steam generating tubes[J]. Teploenergetika, 10, 49-52(1963).

[18] Жукаускас А А, MA Changwen et al. Convective heat transfer in a heat exchanger[M](1986).

[19] FAN Jie, WU Hexin, GOU Junli et al. Research on fully-implicit numerical algorithm of two-fluid three-field two-phase flow model based on JFNK method[J]. Nuclear Techniques, 45, 040605(2022).

[20] Chen F B, Han Z H. Steady-state thermal fluids analysis for the HTR-PM equilibrium core[J]. International Journal of Advanced Nuclear Reactor Design and Technology, 3, 11-17(2021).

[21] Huang J, Gou J, Mao H H et al. Development of a thermal-hydraulic analysis code for helically coiled once-through steam generator[J]. Nuclear Engineering and Design, 364, 110642(2020).