MSc Nuclear Science and Technology / Course details

Introduction to reactor thermal hydraulics

Power cycles, primary coolant systems and overall arrangement of core and fuel assemblies of typical water, gas and liquid metal cooled reactors.

Overall plant characteristics influenced by thermal hydraulics, the energy production, transfer parameters and the thermal design limits.

Heat transfer by conduction

Fourier’s Law of Conduction, thermal conductivity values in typical reactor fuel and clad materials.

Construction of fuel rods and other fuel element geometries.

Equations for steady state temperature drops across the fuel, gas gap and clad of a cylindrical fuel element.

Equations for steady state temperature drops across the fuel and clad of other non-cylindrical fuel elements.

Evolution of temperatures in a fuel element following loss of heat transfer to the coolant.

Heat transfer by convection

Velocity and temperature profiles in heated channel flow for laminar and turbulent cases.

Reynold’s number, Prandtl number and Nusselt number and their significance.

Bulk coolant temperature in a heated channel.

Forced and natural convection.

Newton’s Law of cooling and the heat transfer coefficient.

Correlations for the Nusselt number in laminar flow and turbulent flow for metallic and non-metallic coolants.

Heat transfer coefficients on the surface of fuel elements for turbulent flow of non-metallic liquids, gases and metallic liquids.

Calculation of fuel element surface temperature at a point in a heated channel.

Boiling heat transfer

Pool boiling description.

Description of forced convective boiling and two phase flow.

Two phase flow regime maps.

Enhancement of heat transfer by nucleate boiling.

Correlations for two-phase heat transfer coefficient.

Critical heat flux, Departure from Nucleate Boiling (DNB) and Dryout (DO).

Calculation of critical heat flux, DNB ratio and Critical Power Ratio.

Fluid flow

Pressure drop in a single phase flow

Single phase friction factor.

Circuit pressure drop, flow rate and pumping.

Pressure drop in a heated gas flow.

Pressure drop in a two phase flow.

Flow instabilities in heated channels.

Thermal hydraulic design

Thermal limits for different types of reactor.

Calculation of axial profiles of coolant bulk temperature, clad surface temperature and fuel centre line temperature.

Axial profiles of DNBR or CPR.

Hot channel factors.

Basis of sub-channel computer codes (eg COBRA)

Steam and gas power cycles

U Tube Steam Generators (UTSGs) and Once Through SGs (OTSGs).

Overall heat transfer behaviour of a simplified SG.

The Rankine steam cycle, efficiency and net specific work.

Improvements to the Rankine cycle, superheating, reheating and feed heating.

The Brayton gas turbine cycle, efficiency, regeneration and intercooling.

The unit aims to:

Describe the thermal hydraulic processes involved in the transfer of power from the core to the secondary systems of a nuclear reactor plant and produce competence in the fundamentals of the calculations associated with these processes.

ILO 1

Derive equations for conductive heat transfer in various nuclear fuel elements and evaluate fuel and clad temperatures in the core of a reactor against appropriate thermal hydraulic criteria by applying them.

ILO 2

Describe and explain single phase and multiphase convective heat transfer in coolant channels in nuclear reactor cores with various fuel element arrangements.  

ILO 3

Estimate heat transfer rates and surface temperatures in nuclear reactor cores by selection and application of appropriate convective heat transfer coefficients.  

ILO 4

Explain single phase and multiphase fluid dynamics in coolant channels in nuclear reactor cores with various fuel element arrangements.

ILO 5

Analyse flow rates and coolant pressures in a nuclear reactor cooling system based on fundamental physics principles and the use of empirical correlations.

ILO 6

Explain typical thermal hydraulic safety criteria applied in the core of a nuclear reactor and describe an overview of the typical safety systems used to ensure the thermal hydraulic safety criteria are not exceeded.  

ILO 7

Explain the phenomenon of critical heat flux as a thermal hydraulic limit and use appropriate methods to evaluate the CHF in a heated channel.

ILO 8

Explain the primary models used in industry thermal hydraulic codes including considerations arising from the numerical solution of the governing equations and use an industry thermal hydraulic code to assess a heated channel against limiting thermal hydraulic criteria.

ILO 9

Describe and explain a range of thermodynamic cycles used for power conversion in light water and gas cooled nuclear.

Pre-course online learning  

Face-to-face lectures using presentation software and white board

Tutorials whereby students work through questions to reinforce and consolidate lecture material and additional stretch questions

Hands-on use of a thermal hydraulic code used by industry to assess a case study nuclear heated channel including interpreting computer code output.

Use spread-sheets to perform thermal hydraulic scoping calculations.

Make appropriate assumptions to build models of engineering systems.

Assimilate complex and copious technical information.

Assignment.

Module available as Distance Learning includes:

Online lecture recordings.

Chat room and forums.

Tutorials whereby students work through questions to reinforce and consolidate lecture material and additional stretch questions.

Opportunity to attend session for hands-on use of a thermal hydraulic code used by industry to assess a case study nuclear heated channel.

Use spread-sheets to perform thermal hydraulic scoping calculations.

Make appropriate assumptions to build models of engineering systems.

Assimilate complex and copious technical information.

Assignment. 

Multiple choice test - 20%

Assignment – Open book long answer style questions which tests students understanding of different methods to estimate key thermal hydraulic parameters and how these come together to demonstrate performance and safety of a reactor plant. - 30%

Examination - 50%

Submitted and marked via Canvas

Course Notes

Power Point Slides

“Introduction to Nuclear Engineering”, Lamarsh & Baretta

“Nuclear Systems I”, Todreas & Kazimi

“Nuclear Power Engineering”, El Wakil

“Thermal Design of  Nuclear Reactors “,  R Winterton

“Convective Boiling and Condensation”, Collier

“Engineering Thermodynamics”, Rogers & Mayhew

“Fluid Mechanics”, Gasiorek & Swaffield

Thermal Hydraulic code manuals