Thermohydraulic Design of Nuclear Reactors
- Department of Nuclear Reactors
The course extends theoretical knowledge from the course Thermohydraulics of nuclear reactors and different thermohydraulics courses and shows its practical application for design of nuclear reactors. Students come to know more about flow and heat transfer in the fuel bundles and different methods of thermohydraulic design of reactor core. In details are explained CFD solution, subchannel analysis and use of system codes for these purposes. Coupling of the mentioned methods and coupling with different calculations are explained too. Theoretical lectures are completed with exercises during which students practice theory on practical tasks which are solved by SW codes: CFD ANSYS group, ALTHAMC12, COBRA SFS and RELAP.
- Syllabus of lectures:
1.Flow and heat transfer in fuel bundles and its possibilities of solving (2 lectures): unique features of flow, physical phenomena of flow and heat transfer in reactor core, methods of solving.
2.CFD (3 lectures): basis of method, geometry creation and mesh generation, settings of computations, results analyses, results sensitivity, method advantages and disadvantages, suitability of method.
3.Subchannel analysis (3 lectures): basis of method, geometry creation, settings of computations, results analyses, results sensitivity, method advantages and disadvantages, suitability of method.
4.System codes (3 lectures): basis of method, nodalisation, settings of computations, results analyses, results sensitivity, method advantages and disadvantages, suitability of method.
5.Coupling of methods and codes (2 lectures).
- Syllabus of tutorials:
1.Task solving by CFD code (4 exercises)
2.Task solving by subchannel analysis (4 exercises)
3.Task solving by system code (4 exercises)
4.Tasks evaluation and coupling (1 exercise)
- Study Objective:
- Study materials:
Lamarsh J.R., Baratta A.J.: Introduction to Nuclear Engineering. London: Pearson, 2017. ISBN 978-0134570051
Michener T.E., Cuta J.M., Rector D.R., Adkins H.E., Jr.: COBRA-SFS: A Thermal-hydraulic Analysis Code for Spent Fuel Storage and Transportation Casks, Cycle 4, PNNL-24841, Richland, 2015
D'Auria F. ed.: Thermal-Hydraulics of Water Cooled Nuclear Reactors, Woodhead Publishing, 2017, ISBN: 9780081006627
Todreas N.E., Kazimi M.S.: Nuclear systems, volume II - Elements of Thermal Hydraulic Design, Tailor and Francis, 2001, ISBN 1-56032-079-6
ANSYS: ANSYS help, on line: https://ansyshelp.ansys.com
The RELAP5 Development Team: RELAP5/MOD3.3 Code Manual - Code Structure, System Models and Solution Methods, NUREG/CR-5535/Rev P4, INEL, 2010
- Further information:
- No time-table has been prepared for this course
- The course is a part of the following study plans: