Publications since 2013

Journal articles

Hong Xu, Aurelian Florin Badea, Xu Cheng. Optimization of the Nodalization of Nuclear System Thermal-Hydraulic Code Applied on Primary Loop Benchmark. Journal of Nuclear Engineering and Radiation Science Vol 8 (2022).

Song, M.Q., Liu, X.J., Cheng, X.. Prediction of critical heat flux (CHF) for the high-pressure region in uniformly heated vertical round tubes. Annals of Nuclear Energy, 158 (2021) 108303.

Song, M.Q., Liu, X.J. Assessment of CHF and post-CHF Heat Transfer Models for High-Pressure Condition. Frontiers in Energy Research, 9 (2021) 782086.

Köckert, L., Badea, A. F., Cheng, X., Yu, D., Klingel, D. Studies on post-dryout heat transfer in R-134a vertical flow. International Journal of Advanced Nuclear Reactor Design and Technology 3 (2021) 44–53. DOI: 10.1016/j.jandt.2021.05.001.

Hong Xu, Aurelian Florin Badea, Xu Cheng. Development of a new full-range critical flow model based on nonhomogeneous non-equilibrium model. Annals of Nuclear Energy 158 (2021) 108286.

Hong Xu, Aurelian Florin Badea, Xu Cheng. Studies on the criterion for choking process in two-phase flow. Progress in Nuclear Energy 133 (2021) 103640

Hong Xu, Aurelian Florin Badea, Xu Cheng. Analysis of two phase critical flow with a non-equilibrium model. Nuclear Engineering and Design 372 (2021) 110998

Meng Zhao, FangnianWang, Aurelian Florin Badea. Evaluation and development of heat transfer model for supercritical water flowing in 2 × 2 rod bundles with spacer grid. Volume 165, Part B, February 2021, 120702

Kryková, M., Schulenberg, T., Arnoult Růžičková, M., Sáez-Maderuelo, A., Otic, I., Czifrus, S., Cizelj, L., Pavel, G. L. "European Research Program on Supercritical Water-Cooled Reactor." ASME J of Nuclear Rad Sci. April 2021; 7(2): 021301, https://doi.org/10.1115/1.4048901

Xu, H., Badea, Aurelian F., Cheng, X., 2020. Sensitivity analysis of thermal-hydraulic models based on FFTBM-MSM two-layer method for PKL IBLOCA experiment, Annals of Nuclear Energy 147 (2020) 107732

Jure Oder, Iztok Tiselj, Wadim Jäger, Thomas Schaub, Wolfgang Hering, Ivan Otic, Afaque Shams, "Thermal fluctuations in low-Prandtl number fluid flows over a backward facing step", Nuclear Engineering and Design, Volume 359, 2020, ISSN 0029-549

Liu, X.J., Yang, T., Chai, X., Xiong, J.B., Zhang, T.F., Cheng, X., 2020. Preliminary safety analysis for the SCWR-FQT facility, Int. J Energy Res. 2020; 44: 8103–8112.

Cheng, X., Wielenberg, A., Hampel, U., Starflinger, J., Gupta, S., Schaffrath, A. and Weyermann, F., 2020. Summary of 3rd Sino-German symposium on fundamentals of advanced nuclear safety technology, Kerntechnik, 85 (2020) 2

Wang, F.N, Cheng, X., 2020. Extension and validation of aerosol wash-down model on inclined wall, Annals of Nuclear Energy 144 (2020) 107506

Xiong, J.B., Lu C., Qu W.H., Yang, Y.H., Cheng, X., 2020. Assessment of RANS models in predicting mixing flow induced by split-type vanes in rod bundle, Nuclear Engineering and Design 363 (2020) 110615

Liu, X.J., Yang, D.M., Yang, Y., Chai, X., Xiong, J.B., Zhang, T.F., Cheng, X., 2020. Computational fluid dynamics and subchannel analysis of lead–bismuth eutectic-cooled fuel assembly under various blockage conditions. Applied Thermal Engineering. 2020 Jan 5; 164:114419.

Liu, X.J., Wang, Q., He, Z., Chen, K., Cheng, X., 2019. Phenomena identification and ranking table exercise for thorium based molten salt reactor-solid fuel design. Frontiers in Energy.:1-8.

Zhang, T., Xiong, J.B., Liu, X.J., Chai, X., Li, W., Cheng, X., 2019. Conceptual design of an innovative reduced moderation thoriumfueled small modular reactor with heavywater coolant. International Journal of Energy Research, 2019 Nov; 43(14):8286-98.

Liu, X.J., Song, M.Q., Cheng, X., 2019. Current status and challenges of supercritical fluid thermal hydraulics. Nuclear Engineering and Design. 2019 Dec 1;354:110176.

Qu, W., Xiong, J.B., Chen, S., Cheng,  X., 2019. High-fidelity PIV measurement of cross flow in 5× 5 rod bundle with mixing vane grids. Nuclear Engineering and Design. 2019 Apr 1; 344,:131-43.

Qu, W., Wang, Z., Xiong, J.B., Cheng, X., 2019. Experimental study of cross flow and lateral pressure drop in a 5× 5 rod bundle with mixing vane spacer grid. Nuclear Engineering and Design. 2019 Nov 1;353:110209.

Xu, W., Guo, J., Liu, X.J., Xiong, J.B., Chai, X., Zhang, T., Cheng, X., Zeng, W., 2019. Predictions of quench temperature and quench velocity in narrow rectangular channel of novel plate-type reactor, Annals of Nuclear Energy. 2019 Sep 1;131:148-55.

Jin, D., Xiong, J.B., Cheng, X., 2019. Investigation on interphase force modeling for vertical and inclined upward adiabatic bubbly flow. Nuclear Engineering and Design. 2019 Aug 15; 350:43-57.

Wang, F.N., Cheng, X., 2019. Modeling approach of flowing condensate coverage rate on inclined wall for aerosol wash down, Nuclear Engineering and Design 355 (2019) 110349Shen, D.H., Liu, X.J., Cheng, X., 2019. A study on geometric shape factors for turbulent mixing coefficients in rod bundles, Nuclear Engineering and Design (2019) 353

Luo Y, Liu X, Cheng X. IVR-ERVC study of 1700 MW class PWR based on MAAP simulation and coupled analysis, Annals of Nuclear Energy, 2019, 126: 1-9.

 

Cheng, X., Zhao, M., Feuerstein, F., Liu, X.J., 2019. Prediction of heat transfer to supercritical water at different boundary, International Journal of Heat and Mass Transfer 133 (2019) 527-536

 

Jure Oder, Iztok Tiselj, Wadim Jäger, Thomas Schaub, Wolfgang Hering, Ivan Otic, Afaque Shams. Thermal Fluctuations in Low-Prandtl Number Fluid Flows over a Backward Facing Step, Nuclear Engineering and Design, submitted 2019.

 

Yang, Z, Cheng, X., Zheng, X.H., Chen, H-S., 2019. Numerical investigation on heat transfer of the supercritical fluid upward in vertical tube with constant wall temperature, International Journal of Heat and Mass Transfer 128 (2019) 875–884

 

Wang, X., Cheng, X., 2018. Analysis of inter-channel sweeping flow in wire wrapped 19-rod bundle, Nuclear Engineering and Design 333 (2018) 115–121

 

Shen, D.H., Liu, X.J., Cheng, X., 2018. A new turbulent mixing modeling approach for sub-channel analysis code, Annals of Nuclear Energy 121 (2018) 194–202

Cheng, X., Feuerstein, F., Klingel, D., Yu, D.L., Mechanistic prediction of post dryout heat transfer and rewetting, Kerntechnik 83 (2018) 3, Carl Hanser Verlag, München

Xiong, J.B., Qu, W.H., Wu, Z.H., Cheng, X., 2018. PIV measurement of cross flow in a rod bundle assisted by telecentric optics and matched index of refraction, Annals of Nuclear Energy 120 (2018) 540–545

Liu, X.J., Luo, Y.J., Cao, Z., Guo, R., Cheng, X., 2018. Safety research of IVR-ERVC for advanced water cooled reactor, Energy 156 (2018) 458-467

Badea, Aurelian F., Zhao, M., Cheng, X., Feuerstein, F., Liu, X.J., 2018. Consistency considerations on a large databank and wide range heat transfer prediction for supercritical water in circular tubes, Nuclear Engineering and Design 335 (2018) 178–185

M. Zhao, X. Liu, A.F. Badea, F. Feuerstein and X. Cheng. Comparison of heat transfer models with databank of supercritical fluid, Kerntechnik, Volume 83, Issue 3 (2018) 237-239

Yu, D.L., Feuerstein, F., Koeckert, L., Cheng, X., 2018. Analysis and modeling of post-dryout heat transfer in upward vertical flow, Annals of Nuclear Energy 115 (2018) 186–194

Yang, T., Liu, X.J., Cheng, X., 2018. A circumferentially non-uniform fuel model and its application to thermal-hydraulic code, International Journal of Energy Research, 42(1): 188~197, 2018

Cheng, X., Liu, X.J., 2017. Research Challenges of Heat Transfer to Supercritical Fluids, J of Nuclear Eng. and Rad Sci 4(1), 011003

Dan G. Cacuci, Aurelian F. Badea, Madalina C. Badea, James J. Peltz. Efficient computation of operator-type response sensitivities for uncertainty quantification and predictive modeling: illustrative application to a spent nuclear fuel dissolver model” International Journal for Numerical Methods in Fluids 83 (2017) 149-174.

Aurelian F. Badea, Dan G. Cacuci. Predictive uncertainty reduction in coupled neutron-kinetics/thermal hydraulics modeling of the BWR-TT2 benchmark, Nuclear Engineering and Design 313 (2017) 330–344.

Pang, B., Cheng, X., 2017. Assessment of two-phase interchannel mixing models for BWR and PWR conditions, Annals of Nuclear Energy 01/2017; 99: 283-291

Feuerstein, F., Coelho Silva, A., Klingel, D., Cheng, X., 2017. Large/scale heat transfer experiments with supercritical R134a flowing upward in a circular tube, atw vol.62 (2017), issue 2, pp.121 – 125, February 2017

Cheng, X., Yang, Y.H., Liu, X.J., 2016. Super-Critical Water Reactor with Mixed Spectrum: Design and Key Technologies, Shanghai Jiao Tong University Press, 2016, ISBN 978-7-313-14195-8

Cheng, X., Yang, Y.H., 2016. Status and Challenges of Nuclear Thermal-Hydraulics Research in China for Water-Cooled Reactors, Nuclear Technology, Vol.196, No.2, November 2016, pp.175 – 186

Huang, X., Cheng, X., Klein-Hessling, W., 2016. Evaluation of Passive Containment Cooling with an Advanced Water Film Model in a Lumped-Parameter Code, Nuclear Technology, Vol.196, No.2, November 2016, pp.248 – 259

Xue Zhou Jin, Bradut-Eugen Ghidersa, Aurelian Florin Badea. HELOKA-HP thermal-hydraulic model validation and calibration, Fusion Engineering and Design 109-111, Part B (2016), 1242-1246.

James J. Peltz, Dan Gabriel Cacuci, Aurelian Florin Badea, Madalina C. Badea. Predictive Modeling Applied to a Spent Fuel Dissolver Model—II: Uncertainty Quantification and Reduction, Nuclear Science and Engineering 183 (2016) 332-346.

Cheng, X., 2016. Numerical simulation of multi-phase phenomena in IVR related processes, Kerntechnik 81 (2016) 2, pp.160-163

Liu, X.J., Cheng, X., 2016. Validation of the ATHLET-SC code by trans-critical data, Kerntechnik 81 (2016) 2, pp.164-166

Zhao, M., Gu, H.Y., Li, H.B., Cheng, X., 2016. Heat transfer of water flowing upward in vertical annuli with spacers at high pressure conditionsof Annals, Nuclear Energy, volume 87, 2016, pp. 209 – 216

Cheng, X., et al. 2015. European Activities on Crosscutting Thermal-hydraulic phenomena for innovative nuclear systems, Nuclear Engineering and Design, 290 (2015), 2-12.

Zimmermann, M., Cheng, X., Otic, Sieber, G., Goodheart, K., 2015. Numerical investigation and modeling of two-phase flow sweeping in rod bundles with mixing vane grid spacers, Annals of Nuclear Energy, 85(2015), 403-417.

Agbodemegbe, V.Y., Cheng, X., Akaho, E.H.K., Allotey, F.K.A., 2015. Correlation for cross-flow resistance coefficient using STAR-CCM+simulation data for flow of water through rod bundle supported byspacer grid with split-type mixing vane, Karlsruhe, Nuclear Engineering and Design 285 (2015) 134-149.

Xiong, J.B., Yang, Y.H., Cheng, X., 2015. Numerical analysis on supercritical water heat transfer in a 2x2 rod bundle, Annals of Nuclear Energy, 85(2015),123-134.

Cao, Z, Liu, X.J., Cheng, X.,2015. A two dimensional approach for temperature distribution in reactor lower head during severe accident, Annals of Nuclear Energy,85 (2015), 467-480.

Jin, Y., Xu, W., Liu, X.,J., et al., 2015. In-and ex-vessel coupled analysis of IVR-ERVC phenomenon for large scale PWR. Annals of Nuclear Energy, 2015, 80: 322-337.

Dan Gabriel Cacuci, Aurelian Florin Badea, Predictive modeling methodology for obtaining optimally predicted results with reduced uncertainties: Illustrative application to a simulated solar collector facility, Solar Energy 119 (2015) 486-506.

Yu, Y.Q., Cheng, X, 2014. Experimental study of water film flow on large vertical and inclined flat plate, Progress in Nuclear Energy, volume 77(2014), 176-186.

Chai, X., Otic, I., Cheng, X., 2014. A new drag force model for the wake acceleration effect and its application to simulation of bubbly flow, Progress in Nuclear Energy 80 (2015), 24-36.

Chai, X., Cheng, X., 2014. Wake acceleration effect on spherical bubbles aligned in-line, Progress in Nuclear Energy 80 (2015), 74-79.

Huang, X., Cheng, X., 2014. Modification and application of water film model in COCOSYS for PWR’s passive containment cooling, Nuclear Engineering and Design 280 (2014), 251-261.

Peng, B., Cheng, X., 2014. Proposal of a new type of two-phase interchannel mixing model for application to subchannel analysis of PWR conditions, Annals of Nuclear Energy 73 (2014) 108-121

Yu, Y.Q., Cheng, X, 2014. Experimental study of water film flow on large vertical and inclined flat plate, Progress in Nuclear Energy, volume 77(2014), 176-186.

D.G. Cacuci, A.F. Badea, M.C. Badea. Thermal solar collector predictive modeling: I. Adjoint sensitivity analysis. Transactions of the American Nuclear Society 111 (2014), 1474-1477.

Guo, R., Kuang, B., Cheng, X., A theoretical CHF model for subcooled flow boiling in a curved channel at low pressure. Annals of Nuclear Energy, 2014, 69:196-202.

Xiong, J.B., Cheng, X., Turbulence modelling for supercritical pressure heat transfer in upward tube flow, Nuclear Engineering and Design, 2014, 270, 249-258.

Zhao, M., Gu, H.Y., Cheng, X., Experimental study on heat transfer of supercritical water flowing downward in circular tubes. Annals of Nuclear Energy, 2014, 63: 339-349.

Liu, X.J., Yang, T., Cheng, X., 2013. Thermal-hydraulic analysis of flow blockage in a supercritical water-cooled fuel bundle with sub-channel code, Annals of Nuclear Energy, Annals of Nuclear Energy 59 (2013) 194–203

Zhou, C., Huber, K., Cheng, X., 2013. Validation of the modified ATHLET code with the natural convection test of the PHENIX reactor, Annals of Nuclear Energy, 2013, 59: 31-46

Cheng, X., 2013. R&D on Nuclear Safety and Severe Accident Mitigation in China, Asia Research Policy, Volume 4, Issue 1, 2013

 

Conferences

Aurelian Florin BADEA, Hong XU, and Xu CHENG. SENSITIVITY ANALYSIS AND MODEL IMPROVEMENT BY USING ATHLET PREDICTIONS ON PKL I2.2 BENCHMARK, The 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19), Brussels, Belgium, March 6 - 11, 2022, 36377.

Allen George, Stephan Kelm, and Hans-Josef Allelein. EFFICIENT CFD MODELING OF BULK CONDENSATION, FOG TRANSPORT AND RE-EVAPORATION FOR APPLICATION TO CONTAINMENT SCALE, The 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19), Brussels, Belgium, March 6 - 11, 2022, 19001.

F. Wiltschko et al., Numerical analysis of turbulent heat transfer under super-critical pressure conditions, 10th international Symposium on SCWRs (ISSCWR-10), Prague, the Czech Republic, March 15-19, 2021.

F. Wiltschko et al., One-dimensional investigation of turbulent heat transfer along corroded rod in vertical channels at supercritical pressure conditions, Proceedings of the European Nuclear Young Generation Forum, ENYGF'21, September 27-30, Tarragona, Spain, 2021.

Xu Cheng, Aurelian F. Badea, Meng Zhao. Application of the ATHLET Code to Supercritical Water Cooled Systems, 50th Annual Meeting on Nuclear Technology (AMNT 2019) Berlin, 7 – 8 May 2019

F. Wang et al. Modeling approach of flowing condensate coverage on inclined wall for aerosol wash down, AMNT 2019, 7-8 May, Berlin, Germany.

L. Köckert, F. Feuerstein, D. Yu, D. Klingel, X. Cheng, “Comparison of Post-Dryout Heat Transfer Measurements in an R-134a Cooled Tube and Selected Correlations”, 50th Annual Meeting on Nuclear Technology (AMNT)– Workshop Young Scientists, Berlin, Deutschland, May 7-8, 2019

Song, M.Q., Cheng, X.,2019. Heat transfer analysis of trans-critical pressure transient. AMNT-2019,May 7-8, 2019,Berlin,Germany.

Aurelian F. Badea, Meng Zhao, Xu Cheng. Heat Transfer Predictions for Supercritical Water Flowing in Different Geometries, NUTHOS-12, Qingdao, China, October 14-18, 2018.

Schenk M. CFD Analysis of Liquid Metal Pumps, German CFD Meeting, München, Deutschland; 6.-7. März 2018.

Yu, D., and Cheng, X. (2018). Verification of Proposed Post-Dryout Heat Transfer Model. 49. Jahrestagung Kerntechnik / 49th Annual Meeting on Nuclear Technology (AMNT 2018), Berlin, May 29-30.

L. Köckert, F. Feuerstein, D. Yu, D. Klingel, X. Cheng, “Experimental Study of Post-Dryout Heat Transfer and Rewetting in an R-134a Cooled Vertical Tube at Comparable Water-Cooled Reactor Pressure Conditions”, 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12), Qingdao, China, October 14-18, 2018

Schenk M. CFD Analysis of Centrifugal Liquid Metal Pumps, 49th Annual Meeting on Nuclear Technology – Workshop Young Scientists, Berlin, Deutschland, 29-30 Mai 2018

Schenk M., Cheng X. CFD Analysis of Wall Shear Stress and Recirculation in Centrifugal Liquid Metal Pumps 12th International Topical Meeting on Reactor Thermal-Hydraulics, Operation, and Safety (NUTHOS-12), Qingdao, China, 14.-18. Oktober 2018

F. Wang et al. Model assessment of condensate droplet motion on inclined containment structure surface, NUTHOS12, 14-18 October, 2018, Qingdao, China.

M. Zhao, A.F. Badea, F. Feuerstein, X. Cheng. Assessment of experimental data of heat transfer to supercritical water in tubes, NURETH-17 conference paper, Sep 3-8, 2017.

James J. Peltz, Madalina C. Badea, Dan G. Cacuci, and Aurelian F. Badea. Predictive Modeling of a Paradigm Spent Fuel Dissolver, 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) San Francisco, CA, April 17-20, 2016

Zhao, M., Yang, Q.R., Gu, H.Y., Li, H.B., Cheng, X., 2015. Experimental study on heat transfer of supercritical water flowing upward in 2 2 rod bundles, 46th Annual Meeting on Nuclear Technology, May 5th-7th, 2015, Berlin, Germany.

Meyer, S., Otic, I., Cheng, X., 2015. Phase-Field Modeling of Binary Eutectic Alloy Solidification with Convection, 16th International Topical Meeting on Nuclear Reactor Thermal-hydraulics (NURETH-16), Chicago, USA, August 30 - September 4

Dan G. Cacuci, James J. Peltz, Aurelian F. Badea, and Madalina C. Badea. Spent Fuel Dissolver Predictive Modeling: I. Application of the Adjoint Sensitivity Analysis Methodology for Nonlinear Systems with Operator-Type Responses, ANS Winter Meeting, Washington, DC, US, November 8-12, 2015.

James J. Peltz, Aurelian F. Badea, Madalina C. Badea and Dan G. Cacuci. Spent Fuel Dissolver Predictive Modeling: II. Uncertainty Reduction via Experimental Data Assimilation, ANS Winter Meeting, Washington, DC, US, November 8-12, 2015.

Sonntag, M., Cheng, X., 2015. CFD Model for Simulation of Subcooled Nucleate Flow Boiling and Coupling with STH-Code for Analysis of IVR Cool Ability, International Congress on Advances in Nuclear Power Plants, ICAPP 2015, Nice, France, May 3-6, 2015

Otic, I., Chai, X., Cheng, X., 2014. Pseudo-transient simulation of turbulent mixing in a rectangular channel, Proceedings of ICONE22, 2014

Otic, I., 2014. One equation subgrid model for turbulent convection, Proceedings of NUTHOS-10, 2014

Badea, A.F., Cacuci, D.G., Badea, M.C., 2014. Thermal Solar Collector Predictive Modeling: II. Data Assimilation, Model Calibration, and Best-Estimate Predictions with Reduced Uncertainties, ANS Meeting, Anheim, US, November 9-13, 2014.

Badea, A.F., Cacuci, D.G., Badea, M.C., 2013. Uncertainty Reduction in Calibrated FLICA4 Thermal-Hydraulics Computational Predictions Following Assimilation of Multiple BFBT Benchmark Experimental Data, The 15th International Topical Meeting on Nuclear Reactor Thermalhydraulics, NURETH-15, Pisa, Italy, May 12-15, 2013

Badea, A.F., Cacuci, D.G., Badea, M.C., 2013. Uncertainty Reduction in Coupled Neutron-Kinetics / Thermal-Hydraulics Computational Predictions Following Assimilation of BWR-TT Benchmark Experimental Data, The 15th International Topical Meeting on Nuclear Reactor Thermalhydraulics, NURETH-15, Pisa, Italy, May 12-15, 2013