Thermohydraulics of Nuclear Reactors
Code  Completion  Credits  Range  Language 

17THYR  Z,ZK  4  3P+1C  Czech 
 Vztahy:
 In order to register for the course 17THAR, the student must have received credit for the course 17THYR in a previous semester.
 Garant předmětu:
 Dušan Kobylka
 Lecturer:
 Dušan Kobylka
 Tutor:
 Dušan Kobylka
 Supervisor:
 Department of Nuclear Reactors
 Synopsis:

The course extend student´s basic knowledge in the field of thermohydraulics of nuclear reactors, which they obtain in their previous study. Students are familiarized with 2 phase flow, boiling convection together with forced convection and boiling crisis analyses in the nuclear core conditions. The temperature distribution in the coolant channel will be described together with the thermohydraulic of the full nuclear reactor core, including the hot channel theory. The parts of the course are also lectures about compressible fluid flow theory (ideal gases, vapors, ….) and turbulent flow and its modelling. Explication is focused on understanding and application of knowledge for basic thermohydraulic design of nuclear devices and safety analyses and shows todays limits of knowledge. One lecture is focused on special convection to uncommon coolants, which can be applied for example in gen. IV nuclear reactors.
 Requirements:
 Syllabus of lectures:

1.2 phase flow (3 lectures): theory of 2 phase flow and 2 phase flow calculations, driftflux model and different models for slip calculations, void fraction, pressure losses in 2 phase flow, instabilities of 2phase flows.
2.Nonadiabatic 2 phase flow in heated channel (3 lectures): coolant heating in channel and levels of heating, void fraction and heat transfer calculations in 2 phase flow, void fraction calculations and heat transfer calculations, pressure losses, boiling crisis in the coolant channel in the reactor core.
3.Thermohydraulics of: coolant channel, core and primary circuit (2 lectures): axial and radial temperature distribution in the coolant channel, possibilities of thermohydraulic design of nuclear core, hot channel theory, limits of thermal power in the reactor core.
4.Compressible fluid mechanics (2 lectures): basics of compressible fluid dynamics, critical flow of ideal gas, critical 2 phase flow.
5.Turbulent flow (2 lectures): theoretical possibilities of turbulent flow description, overview and classification of turbulent flow models and their theory.
6.Special convection (1 lecture): coolant comparison, convection to liquid metals, convection to gases, convection to supercritical water and convection in molten salts.
 Syllabus of tutorials:

Selected chapters are completed with practical tasks calculations: type of 2 phase flow determination, 2 phase flow pressure losses, coolant channel heating, critical heat flux, critical flow, Laval nozzle, shock wave.
 Study Objective:
 Study materials:

Key references:
[1]D'Auria F. ed.: ThermalHydraulics of Water Cooled Nuclear Reactors, Woodhead Publishing, 2017, ISBN: 9780081006627
[2]Todreas N. E., Kazimi M. S.: Nuclear systems, volume I  Thermal Hydraulic Fundamentals, CRC Press, 2012, ISBN 9781439808870
Recommended references:
[3]Todreas N.E., Kazimi M.S.: Nuclear systems, volume II  Elements of Thermal Hydraulic Design, Tailor and Francis, 2001, ISBN 1560320796
[4]Wagner W., Kretzschmar H.J.: International Steam Tables, Properties of Water and Steam Based on the Industrial Formulation IAPWSIF97, Second Edition, Springer, 2008, ISBN: 9783540214199
 Note:
 Timetable for winter semester 2024/2025:
 Timetable is not available yet
 Timetable for summer semester 2024/2025:
 Timetable is not available yet
 The course is a part of the following study plans: