Nuclear thermal hydraulics

The objectives of our research are to develop a better understanding of flow and heat transfer phenomena in nuclear reactor systems and to improve their prediction capability.

Power station

We tackle problems encountered in reactors that are currently under operation, as well as performing fundamental research towards the design of new generation reactors. Both RANS and DNS are used.

A focus of our recent fundamental research is the study into mixed convection. Effects of fluid properties, buoyancy and flow acceleration have been studied extensively. Such effects become particularly significant for fluids at supercritical pressure, a primary operating condition in the Supercritical Water-cooled Reactor (SCWR) which is one of the Generation IV candidate reactors.

In dealing with the issues encountered by the industry, we have been studying the flow induced vibration related to Pin-Brace interaction in AGRs, effect of reactor core cross flow and modelling of flow in fuel bundles. We also study instability issues in reactors.

Our key research areas, recent projects and publications are shown below.

Advanced Gas-cooled Reactors (AGRs)

There are currently 15 reactors in operation in the UK and 14 of them are AGRs, altegther producing around a quarter of the electricity used in the country. We perform thermal hydraulics analysis to support the operation and safety case devlopment, including lifletime extension, for such reactors. Our work extends both the reactor core heat transfer and fuel route thermal hydraulics.

  1. Computational modelling for fuel route thermal hydraulics analysis for AGRs, PhD project funded by EPSRC CASE studentship with EDF Energy, 2015-2018
  2. Computational modelling for dropped fuel thermal hydraulics analysis for AGRs, PhD project funded by EDF Energy and University of Sheffield, 2016-2020
  3. Investigation into effect of fuel channel/stringer eccentricity in aged AGRs, PhD project funded by EDF Energy, 2016-2019
  4. Investigation into the complex flow in the fuel stringer and pin-brace interactions within a nuclear reactor, PhD project, Kristin Newlands, 2010-2014.
  5. Studies of cross flow and flow instability in AGRs, PhD project funded by EDF Energy, Yu Duan, 2010-2015

Advanced 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 alternative design concepts are currently being reserached. These are the so-called Gen IV recators. We have been conducting research in support of two of the designs, namely, supercritical water-cooled reactor (SCWR) and sodium-cooled fast reactor (SFR).

  1. Characterizing flow & heat transfer of supercritical fluid in support of design of SCWRs through numerical experiments using DNS, funded by EPSRC through 'Thermal hydraulics for boiling and passive systems (EP/K007777/1)' 2012-2014 and ' Grace Time (EP/M018733/1)' 2015-2019 with Simon Walker (Imperial, PI) and Michael Fairweather (Leeds).
  2. Study of fluid-to-fluid scale for heat transfer to fluids at supercritical pressure using DNS, collaborative research with Universities of Pisa and Tsinghua, 2015-.
  3. Sub-channel mixing & coefficient in fuel bundles for water at supercritical pressure (SCWR), PhD studentship funded by Malaysian Government, 2014-2018
  4. Modelling of turbulent heat transfer to fluids at supercritical pressure, EPSRC (GR/S19424/01), 2003-2006.
  5. Participant of IAEA Coordinated Research Project (CRP) in supercritical water-cooled reactor (SCWR), 2014-2018.
  6. Modelling of heat transfer to sodium in support of design of sodium-cooled fast reactors (SFRs), 2017-2019
  7. Smart models for fuel channel and the pool for SFRs, 2017-2019

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
  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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