[1] Islam S, Haque J M. A comparative study and design optimization of potential cladding materials for gas cooled fast reactor: Neutronics, radiation shielding, and thermal hydraulics analysis[J]. Progress in Nuclear Energy, 185, 105741(2025).

[2] ZHOU Xuanqing, WANG Chen. Research and application on precision control of spherical fuel elements for high-temperature gas-cooled reactors[J]. Enterprise Management, 300-301(2023).

[3] ZHOU Zhenhua, PAN Deng, CHEN Yushuang et al. Core flow distribution design of molten salt reactor with liquid fuel[J]. Nuclear Techniques, 39, 050601(2016).

[4] Bardet P, Blandford E, Fratoni M et al. Design, analysis and development of the modular PB-AHTR[R](2008).

[5] Xiong T, Yan X, Xiao Z J et al. Experimental study on flow instability in parallel channels with supercritical water[J]. Annals of Nuclear Energy, 48, 60-67(2012).

[6] Zheng H, Yu T, Qu C T et al. Basic characteristics and application progress of supercritical water[J]. IOP Conference Series: Earth and Environmental Science, 555, 012036(2020).

[7] ZHANG Shuai. Study on pitch optimization of filament winding for CEFR fuel assembly[D](2024).

[8] SU Shaomin. Joint cross-scale simulation study of the main cooling system of a sodium-cooled fast reactor[D](2023).

[9] Watanabe O, Oyama K, Endo J et al. Development of an evaluation methodology for the natural circulation decay heat removal system in a sodium cooled fast reactor[J]. Journal of Nuclear Science and Technology, 52, 1102-1121(2015).

[10] ZHOU Lei, ZHANG Dalin, LIU Yapeng et al. Three-dimensional numerical simulation of loss of flow accident for fast flux test facility[J]. Atomic Energy Science and Technology, 56, 475-484(2022).

[11] Velusamy K, Natesan K, Maity R K et al. Primary system thermal hydraulics of future Indian fast reactors[J]. Nuclear Engineering and Design, 294, 170-182(2015).

[12] ZHANG Yuhao, ZHAO Haiqi, SHEN Xihao et al. Review on three-dimensional numerical simulation research of flow and heat transfer characteristics in complex space of pool-type sodium-cooled fast reactor[J]. Atomic Energy Science and Technology, 58, 271-283(2024).

[13] Zhao H Q, Lu D G, Zhang Y H et al. Numerical simulation of natural circulation characteristics under different DHX layout schemes in pool-type SFR during station blackout accident[J]. Progress in Nuclear Energy, 157, 104567(2023).

[14] YANG Hongyi, XUE Xiuli, ZHOU Zhiwei et al. Research progress in key thermal-hydraulic issue of sodium-cooled fast ractor[J]. Atomic Energy Science and Technology, 58, 1797-1816(2024).

[15] Muramatsu T, Ninokata H. Investigation of turbulence modelling in thermal stratification analysis[J]. Nuclear Engineering and Design, 150, 81-93(1994).

[16] Ohno S, Ohki H, Sugahara A et al. Validation of a computational simulation method for evaluating thermal stratification in the reactor vessel upper plenum of fast reactors[J]. Journal of Nuclear Science and Technology, 48, 205-214(2011).

[17] Tenchine D, Vandroux S, Barthel V et al. Experimental and numerical studies on mixing jets for sodium cooled fast reactors[J]. Nuclear Engineering and Design, 263, 263-272(2013).

[18] Yu Y Q, Shemon E R, Kim T K et al. Evaluation of hot channel factor for sodium-cooled fast reactors with multi-physics toolkit[J]. Nuclear Engineering and Design, 365, 110704(2020).

[19] Uitslag-Doolaard H J, Alcaro F, Roelofs F et al. Multiscale modelling of the phénix dissymmetric test benchmark[J]. Nuclear Engineering and Design, 356, 110375(2020).

[20] Bavière R, Tauveron N, Perdu F et al. A first system/CFD coupled simulation of a complete nuclear reactor transient using CATHARE2 and TRIO_U. Preliminary validation on the Phénix Reactor Natural Circulation Test[J]. Nuclear Engineering and Design, 277, 124-137(2014).

[21] Zhao H Q, Zhang Y H, Pu Q F et al. Analysis on thermal hydraulic characteristics for PHENIX natural circulation system using "1D system code + 3D CFD" simulation method[J]. Progress in Nuclear Energy, 176, 105370(2024).

[22] Shams A, Roelofs F, Baglietto E et al. High fidelity numerical simulations of an infinite wire-wrapped fuel assembly[J]. Nuclear Engineering and Design, 335, 441-459(2018).

[23] Angeli D, Fregni A, Stalio E. Direct numerical simulation of turbulent forced and mixed convection of LBE in a bundle of heated rods with P/D=1.4[J]. Nuclear Engineering and Design, 355, 110320(2019).

[24] Cheng X, Tak N I. Investigation on turbulent heat transfer to lead–bismuth eutectic flows in circular tubes for nuclear applications[J]. Nuclear Engineering and Design, 236, 385-393(2006).

[25] Johnson H A, Hartnett J P, Clabaugh W J. Heat transfer to molten lead-bismuth eutectic in turbulent pipe flow[J]. Journal of Fluids Engineering, 75, 1191-1198(1953).

[26] Kirillov P L, Ushakov P A. Heat transfer to liquid metals: specific features, methods of investigating, and main relationships[J]. Thermal Engineering, 48, 50-59(2001).

[27] DENG Shiyu, LU Tao, DENG Jian et al. Optimization of turbulent prandtl numbers and rans models for liquid lead-bismuth eutectic[J]. Nuclear Power Engineering, 44, 98-103(2023).

[28] Hwang C B, Lin C A. A low Reynolds number two-equation kθ-ϵ˜θ model to predict thermal fields[J]. International Journal of Heat and Mass Transfer, 42, 3217-3230(1999).

[29] Nagano Y, Shimada M. Development of a two-equation heat transfer model based on direct simulations of turbulent flows with different Prandtl numbers[J]. Physics of Fluids, 8, 3379-3402(1996).

[30] Manservisi S, Menghini F. Triangular rod bundle simulations of a CFD k-ϵ-kθ-ϵθ heat transfer turbulence model for heavy liquid metals[J]. Nuclear Engineering and Design, 273, 251-270(2014).

[31] Fiore M, Koloszar L, Fare C et al. Physics-constrained machine learning for thermal turbulence modelling at low Prandtl numbers[J]. International Journal of Heat and Mass Transfer, 194, 122998(2022).

[32] Kays W M. Turbulent Prandtl number: where are we[J]. Journal of Heat Transfer, 116, 284-295(1994).

[33] MA Rongfang, JIANG Gefan, FANG Yonghan et al. The Molten Salt Reactor (MSR): overview and perspectives[J]. Nuclear Science and Engineering, 44, 1239-1246(2024).

[34] Ingersoll D T, Forsberg C W, Ott L J et al[M]. Status of preconceptual design of the advanced high-temperature reactor (AHTR)(2004).

[35] Meng X K, Sun Z N, Xu G Z. Single-phase convection heat transfer characteristics of pebble-bed channels with internal heat generation[J]. Nuclear Engineering and Design, 252, 121-127(2012).

[36] PAN Deng, YU Xiaohan, ZOU Yang et al. CFD-simulation of structured packed beds molten salt reactors on pressure drop and heat transfer[J]. Nuclear Techniques, 39, 080604(2016).

[37] JING Jianping, WU Lin, XIA Yuqi et al. Study on characteristic of flow and heat transfer in 10-MWt molten salt reactor-solid fuel[J]. Nuclear Science and Engineering, 38, 135-140(2018).

[38] Bao J[J]. Development of the model for the multi-physics analysis of Molten Salt Reactor Experiment using GeN-Foam code(2016).

[39] GE Jian, ZHANG Dalin, TIAN Wenxi et al. Coupled neutronics and thermal hydraulics analysis for molten salt reactor based on CFD method[C], 734-738(2015).

[40] Sohal M S, Ebner M A, Sabharwall P et al. Engineering database of liquid salt thermophysical and thermochemical properties[R](2010).

[42] Podila K, Chen Q, Rao Y F. CFD simulations of molten salt reactor experiment core[J]. Nuclear Science and Engineering, 193, 1379-1393(2019).

[43] HE Fan. Thermal-hydraulic transient analysis of molten salt reactor based on RELAP5/FLUENT coupling program[D](2021).

[44] HE Fan, CAI Xiangzhou, GUO Wei et al. Development of the RELAP5/FLUENT coupled program[J]. Atomic Energy Science and Technology, 55, 693-703(2021).

[45] Hu T L, Cao L Z, Wu H C et al. Coupled neutronics and thermal-hydraulics simulation of molten salt reactors based on OpenMC/TANSY[J]. Annals of Nuclear Energy, 109, 260-276(2017).

[46] HU Chuanwei, Yanzhi E, ZOU Yang et al. Research on GPU parallelization of fluid dynamics process of molten salt reactor core[J]. Nuclear Techniques, 40, 61-67(2017).

[47] CHENG Xu, LIU Xiaojing. Research status and prospect of supercritical water-cooled reactor[J]. Atomic Energy Science and Technology, 42, 167-172(2008).

[48] Cheng X, Kuang B, Yang Y H. Numerical analysis of heat transfer in supercritical water cooled flow channels[J]. Nuclear Engineering and Design, 237, 240-252(2007).

[49] Gu H Y, Cheng X, Yang Y H. CFD analysis of thermal–hydraulic behavior in SCWR typical flow channels[J]. Nuclear Engineering and Design, 238, 3348-3359(2008).

[50] XU Duoting, LI Hongbo, YANG Jue et al. Numerical simulation of heat transfer to SCW in 4-rod bundle[J]. Atomic Energy Science and Technology, 49, 292-296(2015).

[51] MA Dongliang, ZHOU Tao, HUANG Yanping. Computational analysis of asymmetric heating in parallel channels with supercritical water[J]. Nuclear Science and Engineering, 42, 888-895(2022).

[52] Li J J, Zhou T, Song M Q et al. CFD analysis of supercritical water flow instability in parallel channels[J]. International Journal of Heat and Mass Transfer, 86, 923-929(2015).

[53] Xi X, Xiao Z J, Yan X et al. Numerical simulation of the flow instability between two heated parallel channels with supercritical water[J]. Annals of Nuclear Energy, 64, 57-66(2014).

[54] Podila K, Rao Y F. CFD analysis of flow and heat transfer in Canadian supercritical water reactor bundle[J]. Annals of Nuclear Energy, 75, 1-10(2015).

[55] Ebrahimnia E, Chatoorgoon V, Ormiston S J. Numerical stability analyses of upward flow of supercritical water in a vertical pipe[J]. International Journal of Heat and Mass Transfer, 97, 828-841(2016).

[56] HUANG Yanping, ZANG Jinguang. Overview of gas-cooled fast reactor[J]. Modern Physics, 30, 40-43(2018).

[57] Angelo G, Andrade D A, Angelo E et al. A three-dimensional thermal and fluid dynamics analysis of a gas cooled subcritical fast reactor driven by a D-T fusion neutron source[J]. Annals of Nuclear Energy, 38, 2734-2741(2011).

[58] Orosz G I, Aszódi A. CFD modelling of mixing vane spacer grids for ALLEGRO relevant gas cooled reactor fuel geometry[J]. Annals of Nuclear Energy, 164, 108628(2021).

[59] Berthoux M, Cadiou T. The thermal hydraulics in a rod bundle representative of the start-up core of the ALLEGRO Gas cooled Fast Reactor: experimental and numerical approaches[J]. Nuclear Engineering and Design, 240, 3372-3386(2010).

[60] Orosz G I, Magyar B, Szerbák D et al. ALLEGRO Gas cooled Fast Reactor Rod Bundle investigations with CFD and PIV method[J]. Nuclear Engineering and Design, 400, 112062(2022).

[61] Ding C, Kvizda B, Dubey A et al. CFD simulation of injection point design for emergency core cooling system of ALLEGRO[J]. Annals of Nuclear Energy, 213, 111084(2025).

[62] Kvizda B, Mayer G, Vácha P et al. ALLEGRO Gas-cooled Fast Reactor (GFR) demonstrator thermal hydraulic benchmark[J]. Nuclear Engineering and Design, 345, 47-61(2019).

[63] Siwiec A, Spirzewski M. Thermal-hydraulic analysis of GFR demonstrator-Allegro[C](2018).

[64] Vácha P, Bělovský L. Analyses of severe accidents in GFR demonstrator ALLEGRO with MELCOR 2.1[C](2016).

[65] Mayer G, Bentivoglio F. Preliminary study of the decay heat removal strategy for the gas demonstrator ALLEGRO[J]. Nuclear Engineering and Design, 286, 67-76(2015).

[66] Petti D, Shen W, Tuzov A et al. Technology roadmap update for generation IV nuclear energy systems[R](2014).

[67] YAN An, SUN Ximing, DONG Yujie. Simulation study of gas-solid two-phase flow coupling of pebble-bed cores in a high-temperature reactor[J]. Journal of Harbin Engineering University, 43, 1743-1749(2022).

[68] DONG Jianhua, WANG Jun, GUO Juanjuan et al. Simulation on gap bypass flow of the prismatic high temperature gas cooled reactor core[J]. Nuclear Techniques, 44, 120601(2021).

[69] Wu C Y, Ferng Y M, Chieng C C et al. Investigating the advantages and disadvantages of realistic approach and porous approach for closely packed pebbles in CFD simulation[J]. Nuclear Engineering and Design, 240, 1151-1159(2010).

[70] ZHANG Shuangbao, LI Liangxing, XIE Wei et al. Three-dimensional modeling of pebble bed type high temperaturegas-cooled reactor core and steady-state thermal hydraulic analysis[J]. Journal of the University of Chinese Academy of Sciences, 37, 186-191(2020).

[71] PENG Haoran, SUN Jun. Investigation of flow and heat transfer in the pebble bed of the high temperature gas-cooled reactor[J]. Journal of Engineering Thermophysics, 37, 1076-1083(2016).

[72] Gámez Rodríguez A, Rojas Mazaira L Y, García Hernández C R et al. An integral 3D full-scale steady-state thermohydraulic calculation of the high temperature pebble bed gas-cooled reactor HTR-10[J]. Nuclear Engineering and Design, 373, 111011(2021).

[73] Lee J J, Park G C, Kim K Y et al. Numerical treatment of pebble contact in the flow and heat transfer analysis of a pebble bed reactor core[J]. Nuclear Engineering and Design, 237, 2183-2196(2007).

[74] LIU Shixian, LI Xiaowei, WU Xinxin. Numerical simulation of the coolant flow and heat transfer characteristics in the reactor core of pebble bed HTGR[J]. Journal of Engineering Thermophysics, 34, 2106-2111(2013).

[75] LU Tao, LI Yang. Numerical simulation of flow and heat transfer of high temperature gas cooled reactor with body centered cubic fuel sphere packed bed[J]. Journal of Thermal Science and Technology, 15, 273-278(2016).

[76] Wang H H, Dominguez-Ontiveros E, Hassan Y A. Computational fluid dynamics analysis of core bypass flow and crossflow in a prismatic very high temperature gas-cooled nuclear reactor based on a two-layer block model[J]. Nuclear Engineering and Design, 268, 64-76(2014).

[77] Ham T K, Arcilesi D J, Kim I H et al. Computational fluid dynamics analysis of the initial stages of a VHTR air-ingress accident using a scaled-down model[J]. Nuclear Engineering and Design, 300, 517-529(2016).

[78] Ferng Y M, Chi C W. CFD investigating the air ingress accident occurred in a HTGR_simulation of thermal-hydraulic characteristics[J]. Nuclear Engineering and Design, 245, 28-38(2012).

[79] Oh C H, Kim E S, Kang H S. Natural circulation patterns in the VHTR air-ingress accident and related issues[J]. Nuclear Engineering and Design, 249, 228-236(2012).

[80] Ferng Y M, Chi C W. CFD investigating the air ingress accident for a HTGR simulation of graphite corrosion oxidation[J]. Nuclear Engineering and Design, 248, 55-65(2012).

[81] Williamson R L, Hales J D, Novascone S R et al. Multidimensional multiphysics simulation of nuclear fuel behavior[J]. Journal of Nuclear Materials, 423, 149-163(2012).

[82] Liu R, Prudil A, Zhou W Z et al. Multiphysics coupled modeling of light water reactor fuel performance[J]. Progress in Nuclear Energy, 91, 38-48(2016).

[83] Wang Z, Chungyu Z, Yuan C X. Multiphysics modeling of pressurized water reactor fuel performance[J]. Journal of Nuclear Engineering and Radiation Science, 4, 031008(2018).

[84] Brown P N, Saad Y. Hybrid Krylov methods for nonlinear systems of equations[J]. SIAM Journal on Scientific and Statistical Computing, 11, 450-481(1990).

[85] Balay S, Abhyankar S, Adams M et al[M]. PETSc users manual(2019).

[86] Zou L, Hu G J, O'Grady D et al. Explicit modeling of pebble temperature in the porous-media model for pebble-bed reactors[J]. Progress in Nuclear Energy, 146, 104175(2022).

[87] Tano M, Kyriakopoulos V, Arment M C. Validation of Pronghorn's subchannel code using EBR-Ⅱ shutdown heat removal tests: SHRT-17 and SHRT-45R[J]. Nuclear Engineering and Design, 416, 112783(2024).

[88] Wu Y J, Liu B K, Zhang H et al. A multi-level nonlinear elimination-based JFNK method for multi-scale multi-physics coupling problem in pebble-bed HTR[J]. Annals of Nuclear Energy, 176, 109281(2022).

[89] Wu X, Kozlowski T, Meidani H. Kriging-based inverse uncertainty quantification of nuclear fuel performance code BISON fission gas release model using time series measurement data[J]. Reliability Engineering & System Safety, 169, 422-436(2018).

[90] YUE Mingkai, JIN Hao, LIU Tongwei et al. Thermal-mechanical coupling analysis and structure optimization of solid core of heat-pipe-cooled reactor[J]. Atomic Energy Science and Technology, 57, 428-437(2023).

[91] Merzari E, Bolotnov I, Dinh N et al. Building a multiscale framework: an overview of the NEAMS thermal-hydraulics integrated research project[J]. Nuclear Engineering and Design, 433, 113807(2025).

[92] Alemberti A, Carlsson J, Malambu E et al. European lead fast reactor: ELSY[J]. Nuclear Engineering and Design, 241, 3470-3480(2011).

[93] Kefeng LYU. Study on thermal and hydraulic behavior of liquid lead-bismuth alloy in bundle assembly with wires[D](2016).

[94] Bandini G, Polidori M, Gerschenfeld A et al. Assessment of systems codes and their coupling with CFD codes in thermal-hydraulic applications to innovative reactors[J]. Nuclear Engineering and Design, 281, 22-38(2015).

[95] Trucano T, Post D. Guest editors' introduction: verification and validation in computational science and engineering[J]. Computing in Science & Engineering, 6, 8-9(2004).

[96] Slotnick J, Khodadoust A, Alonso J et al. CFD Vision 2030 study: a path to revolutionary computational aerosciences[R](2014).

[97] ZHANG Xiaoyu. Application of polynomial chaos method in uncertainty analysis of core calculation[D](2019).

[98] Xiu D B, Karniadakis G E. The Wiener-askey polynomial chaos for stochastic differential equations[J]. SIAM Journal on Scientific Computing, 24, 619-644(2002).

[99] Lacor C, Dinescu C, Hirsch C et al. Implementation of intrusive polynomial chaos in CFD codes and application to 3D navier-stokes[M]. Uncertainty Quantification in Computational Fluid Dynamics, 193-223(2013).

[100] Knio O M, Le Maître O P. Uncertainty propagation in CFD using polynomial chaos decomposition[J]. Fluid Dynamics Research, 38, 616-640(2006).

[101] Mathelin L, Hussaini M Y, Zang T A. Stochastic approaches to uncertainty quantification in CFD simulations[J]. Numerical Algorithms, 38, 209-236(2005).

[102] WANG Xiaodong, KANG Shun. Application of polynomial chaos method in simulation of random square cavity flow[J]. Scientia Sinica (Technologica), 41, 790-798(2011).

[103] Onorato G, Loeven G, Ghorbaniasl G et al. Comparison of intrusive and non-intrusive polynomial chaos methods for CFD applications in aeronautics[C](2010).

[104] Gopinathrao N P, Bagshaw D, Mabilat C et al. Non-deterministic CFD simulation of a transonic compressor rotor[C], 1125-1134(2010).

[105] Hosder S, Walters R. Non-intrusive polynomial chaos methods for uncertainty quantification in fluid dynamics[C], AIAA2010-129(2010).

[106] Babcsány B, Giusti V, Moise A et al. Results and lessons learned from the Generation IV SCWR-FQT comprehensive Monte Carlo computational benchmark[J]. Annals of Nuclear Energy, 191, 109903(2023).

[108] Micaelli J C[R]. CSNI international standard problem procedures-CSNI Report No.17 - Revision 4(2004).

[109] Mascari F, Bersano A, Alcaro F et al. OECD/NEA/CSNI/WGAMA PERSEO benchmark: main outcomes and conclusions[J]. Nuclear Engineering and Design, 405, 112220(2023).

[110] Lübbesmeyer D, Aksan S N. ISP42 (PANDA Tests)-blind phase comparison report[R](2003).

[111] ZHANG Xueyan, YANG Jun, WANG Shiqi et al. Stage progress and analysis of international standard problem 51 "Open Test"[J]. Nuclear Power Engineering, 43, 15-23(2022).

[113] Wu J, Su X K, Gong Z Q et al. Exploration of high-efficiency heat transfer modes in lead-cooled assemblies: a structural optimization based on the SST k-ω-kθ-ɛθ model[J]. Applied Thermal Engineering, 247, 123051(2024).

[114] Dovizio D, Mikuž B, Shams A et al. Validating RANS to predict the flow behavior in wire-wrapped fuel assemblies[J]. Nuclear Engineering and Design, 356, 110376(2020).

[115] Wu J, Tang Z H, Cai J J. Thermal-hydraulic analysis of LBE flow in rod bundles with wire spacers based on heat transfer framework of OpenFOAM[J]. Nuclear Engineering and Design, 417, 112843(2024).

[116] Shams A, De Santis A, Roelofs F. An overview of the AHFM-NRG formulations for the accurate prediction of turbulent flow and heat transfer in low-Prandtl number flows[J]. Nuclear Engineering and Design, 355, 110342(2019).

[117] Manceau R. Recent progress in the development of the elliptic blending Reynolds-stress model[J]. International Journal of Heat and Fluid Flow, 51, 195-220(2015).

[118] Dong K J, Khan S A, He S H et al. Bubble transport during SGTR accident in lead-cooled fast reactor: a machine learning[J]. Nuclear Engineering and Design, 415, 112724(2023).

[119] Ling J, Kurzawski A, Templeton J. Reynolds averaged turbulence modelling using deep neural networks with embedded invariance[J]. Journal of Fluid Mechanics, 807, 155-166(2016).

[120] Fiore M, Koloszar L, Mendez M A et al. Turbulent heat flux modelling in forced convection flows using artificial neural networks[J]. Nuclear Engineering and Design, 399, 112005(2022).

[121] Chen L X, Yuan C, Zhang H N et al. Development of turbulent heat flux model for unsteady forced convective heat transfer of small-to-medium Prandtl-number fluids based on deep learning[J]. International Journal of Heat and Mass Transfer, 194, 123115(2022).

[122] XIA Chao, ZHOU Hua, YANG Zhigang. Augmentation of Turbulence models using Field Inversion and Machine Learning Applied to Numerical Simulation of Vehicle Aerodynamics[C], 10(2022).

[123] FANG Peixun, HE Chuangxin, XU Sihua et al. Calibration of turbulence model constants using measurement data assimilation: prediction of steam valve flow characteristics with filter[J]. Acta Aerodynamica Sinica, 39, 12-22(2021).

[124] LI Yi, QU Wenhai, XIONG Jinbiao. Turbulence model constant calibration for flow field in rod bundle based on data assimilation[J]. Atomic Energy Science and Technology, 58, 1413-1423, 1387(2024).

[125] Liu Y, Hu R, Kraus A et al. Data-driven modeling of coarse mesh turbulence for reactor transient analysis using convolutional recurrent neural networks[J]. Nuclear Engineering and Design, 390, 111716(2022).

[126] Yang T, Zhao P C, Zhao Y N et al. Development of reduced-order thermal stratification model for upper plenum of a lead–bismuth fast reactor based on CFD[J]. Nuclear Engineering and Technology, 55, 2835-2843(2023).

[127] LIN Ming, CHENG Maosong, CAI Xiangzhou et al. Applicability analysis of reduced order modeling methods for fluid dynamics in molten salt reactor[J]. Nuclear Techniques, 47, 090604(2024).

[128] Li J C, Qian H, Chen G L et al. Research on fine analysis and accelerated prediction technology for thermal stratification in the upper plenum of the "Monju" reactor[J]. Case Studies in Thermal Engineering, 60, 104844(2024).

[129] QIAN Hao, CHEN Guangliang, LIU Dong et al. Deep learning solution technology for sparse data of multi-channel flow field of PWR rod bundle[J]. Nuclear Power Engineering, 46, 81-89(2025).

[130] Su S M, Chen G L, Wu H P et al. Research of PM-ADC scheme for optimized computational CFD analysis in fast reactor cores[J]. International Communications in Heat and Mass Transfer, 150, 107201(2024).

[131] ZHANG Yuyang. Study on the effect of positioning mode on thermal and hydraulic characteristics of Lead Cooled Fast Reactor fuel assembly[D](2023).

[132] ZHANG Zuoyi. Core meltdown-free safety features of a high-temperature gas-cooled reactor pebble-bed module[J]. Scientia Sinica (Technologica), 55, 462-474(2025).

[133] Wu J W, Wu R J, Wang Y Q et al. Prediction of dissolved impurities and movement of oxide particles in the primary circuit of LBE fast reactor[J]. Coatings, 11, 1263(2021).

[134] Ellis M, Gaston D, Forget B et al. Preliminary coupling of the Monte Carlo code OpenMC and the multiphysics object-oriented simulation environment for analyzing Doppler feedback in Monte Carlo simulations[J]. Nuclear Science and Engineering, 185, 184-193(2017).

[135] Tenchine D, Baviere R, Bazin P et al. Status of CATHARE code for sodium cooled fast reactors[J]. Nuclear Engineering and Design, 245, 140-152(2012).