Nuclear thermal hydraulics

Our research is aimed at advancing the understanding of flow and heat transfer phenomena in nuclear reactor systems and improving their prediction capabilities to support new-generation reactor design as well as the safety of existing reactors.

Power station

We tackle problems encountered in reactors that are currently under operation, as well as perform fundamental research towards the design of new generation reactors. We develop and apply modelling tools using both conventional CFD (RANS) and high fidelity CFD (DNS/LES).

Regarding Generation IV reactors, our research is directed at Liquid Metal Fast Reactors (LMFRs, e.g., SFR and LFR) to study cover-gas region aerosol dynamics, stratification and stagnation and thermal striping,  Supercritical Water-cooled Reactor (SCWR) to investigate laminarisation and hear transfer deterioration due to buoyancy and strong variations of thermal properties, and High Temperature Gas-Cooled Reactor (HTGR) to study gas-ingress to the reactor core following a pressure boundary breach.

In dealing with the issues encountered in existing AGRs, we analyse the effects of core distortions and cross flows on the brick cooling in the ageing reactor cores. We study flow induced vibration related to pin-brace interactions in AGRs and perform modelling of flows in fuel bundles. We develop 3D modelling tools for the fuel route thermal analysis. 

Our research is funded by the research councils (EPSRC), government (BEIS), EU (H2020), IAEA and the industry (EDF). 

Find out our key research areas, recent projects and publications from below.

Generation IV reactor design concepts

The Generation IV International Forum (GIF) was set up to carry out research and development on the next generation nuclear energy systems. Six design concepts are currently being researched. These are the so-called Gen IV reactors. We have been conducting research to advance the prediction and understanding of the reactor thermal hydraulics in support of three of the designs, namely, supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR) and high temperature gas-cooled reactor (HTGR).

Advanced Gas-cooled Reactors (AGRs)

There are currently 15 reactors in operation in the UK and 14 of them are AGRs, altogether producing around a quarter of the electricity used in the country. We perform thermal hydraulics analysis to support the operation and safety case development, including lifetime extension, for such reactors. Our work extends both the reactor core heat transfer and fuel route thermal hydraulics. Examples include analysis of the the impact of fuel channel distortion on graphite brick temperatures in the ageing reactors and the assessment of cross flows in the core, which are used in support of reactor lifetime extension safety cases. We have also developed a 3D modelling tool, FREEDOM, to support AGR fuel route thermal analysis and safety cases.   

Fundamental and generic research

We conduct numerical experiments using DNS/LES to develop a better understanding of commonly encountered phenomena in nuclear reactors, and in particular covering the following topics (also see research on Turbulence): 

  • Mixed convection and heat transfer deterioration
  • Natural circulation
  • Flow laminarisation
  • Transient flow and transition
  • Flow over rough surfaces
Tools and methodology development

We develop and maintain a number of modelling packages using high fidelity DNS (CHAPSim), or coarse-grid CFD (SubChCFD) or porous-medium methodology (FREEDOM).

  • CHAPSim - a finite-difference based DNS code initially developed in our group, and is now being developed by CCP-NTH as a nuclear community code.
  • SubChCFD - a novel methodology combing coarse-grid CFD and sub-channel approaches, developed as part of BEIS Nuclear Innovation Program in Digital reactor design - thermal hydraulics.
  • FREEDOM - a three-dimensional model for AGR fuel route thermal hydraulics analysis developed with the support of EDF.  
  1. Duan Y & He S (2017) Large eddy simulation of a buoyancy-aided flow in a non-uniform channel – Buoyancy effects on large flow structures. Nuclear Engineering and Design, Volume 312, February 2017, Pages 191–204.
  2. Duan Y, He S, Ganesan P & Gotts J (2014) Analysis of the horizontal flow in the advanced gas-cooled reactor. Nuclear Engineering and Design, 272, 53-64.
  3. Ganesan P, He S, Hamad F & Gotts J (2013) Effect of horizontal cross flow on the heat transfer form the moderator bricks in the Advanced Gas-cooled Reactor: A CFD study. Nuclear Engineering and Design, 263, 151-163.
  4. M. Sharabi, W. Ambrosini, S. He, P.X. Jiang, C. Zhao, ‘Transient Three-Dimensional Stability Analysis of Supercritical Water Reactor Rod Bundle Subchannels by a Computational Fluid Dynamics Code’, Journal of Engineering for Gas Turbines and Power, Vol. 131, DOI:10.1115/1.3032437, 2009 S.
  5. He, W.S. Kim and J.H., Bae, Assessment of performance of turbulence models in predicting supercritical pressure heat transfer in a vertical tube, International Journal of Heat and Mass Transfer, Vol. 5, pp4659-4675, 2008.
  6. S. He, W.S. Kim & J.D. Jackson, ‘A computational study of convective heat transfer to carbon dioxide at a pressure just above the critical value’, Applied Thermal Engineering, Vol. 28, pp1662-1675, 2008
  7. M. Sharabi, W. Ambrosini, S. He, J.D. Jackson, ‘Prediction of turbulent convective heat transfer to a fluid at supercritical pressure in square and triangular channels’, Annals of Nuclear Energy, Vol. 35, pp767-782, 2008
  8. M. Bucci, M. B. Sharabi, W. Ambrosini, N. Forgione, F. Oriolo, S. He, ‘Prediction of transpiration effects on heat and mass transfer by different turbulence models’, Nuclear Engineering and Design, Vol. 238, pp958-974, 2008.
  9. M. B. Sharabi, W. Ambrosini, S. He, ‘Prediction of unstable behaviour in a heated channel with water at supercritical pressure by CFD models’, Annals of Nuclear Energy, Vol. 35, pp993-1005, 2008.
  10. S. He, P.X. Jiang, Yi-Jun Xu, Run-Fu Shi , W.S. Kim and J.D. Jackson, ‘A computational study of convection heat transfer to CO2 at supercritical pressure in a vertical mini tube’, International Journal of Thermal Sciences, 2005, vol. 44(6), pp.521-530.
  11. S. He and J. Gotts, ‘A Computational study of the effect of distortions of the moderator cooling channel of the Advanced gas-cooled Reactor’, Nuclear Engineering and Design, vol. 235 (9) pp. 965-982, 2005.
  12. S. He and J. Gotts, ‘Calculation of friction coefficient for non-circular channels’, Journal of Fluids Engineering, vol. 126(6), pp. 1033-1038, 2004.
  13. S. He, W.S. Kim, PX Jiang & J.D. Jackson, ‘Simulations of mixed convection heat transfer to carbon dioxide at supercritical pressure’, Journal of Mechanical Engineering Science, vol. 218, pp. 1281-1296, 2004.

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