The course will provide knowledge of the nuclear power plant features, including innovative and Generation IV reactors, and the design methodologies of the main components and systems of a nuclear plant, with particular reference to the thermal-hydraulics in the core and primary loops of light water reactors.
The course consists of lectures and numerical exercises.

At the end of the course the students should be able to cope with the components calculations and the thermal-hydraulic design of the nuclear power plants core, with particular reference to light water reactors and to the thermal limits in the hot subchannel; they should know the nuclear power plants operational characteristics, with particular reference to the innovative reactors and Generatio IV future reactors as well as the decommissioning options.

Basic knowledge of thermodynamics, reactor physics, nuclear power plants , single-phase and two-phase thermal-fluid-dynamics, and heat transfer.

1. Nuclear reactor design and Regulatory guides; safety and seismic categories; quality assurance. Core thermal hydraulic design methodology.
2. Characteristics of the core of different nuclear reactors: comparison of the core thermal and hydraulic parameters, safety margins.
3. Thermal core design: fuel rod thermal limits; fuel rod behaviour and failure mechanisms.
4. Nuclear reactor heat generation. Hot channel factors. Hot subchannel analysis. Fuel rod temperature profile for axial cosinusoidal profile and with flux peak; bypass flow rate; preliminary LWR core design; orificing, thermal crisis and flow rate choice
5. Thermal-hydraulic design of a PWR core: temperatures in the hot channel: coolant, fuel rod wall, cladding, pellet, temperature profiles, conductivity integral, flux depression factor, nucleate boiling on the wall, gap thermal resistance.
6. PWR boiling crisis: definition and phenomenological description, models and correlations in pool boiling, mechanisms for the forced convection and effect of the main thermal parameters, effect of flux distribution, Departure from nucleate boiling. PWR design correlations.
7. Thermal-hydraulic design in boiling water reactors. Correlations for the flow quality, void fraction evaluation, single and two-phase pressure drops.
8. BWR thermal crisis: dryout mechanisms. Correlations for BWR design.
9. IAEA International accidents scale. Major nuclear accidents,accidents prevention and mitigation: filtered vented containment
11. Design of new reactors. Intrinsic and passive safety: IAEA passive categories. Next generation reactors with passive features: evolutionary and innovative reactors, Generation IV reactors.
12. Decommissioning of nuclear plants.

The theoretical lectures are complemented by numerical evaluations concerning the thermal-hydralics in advanced nuclear reactors components and in the PWR hot channel (with verification of the fuel rod limits)

- B. Panella, Appunti.
- C. Lombardi, Impianti nucleari, Cittą Studi, 2004.
- M. Cumo, Impianti nucleari. Casa Editrice Universitą La Sapienza, 2008.
- R.A.Knief,"Nuclear Engineering", Hemisphere,1992.
- N.E.Todreas and M.S.Kazimi,"Nuclear systems",Vol.I ,II,Hemisphere,1990.
- R.T.Lahey and F.J.Moody,"The thermal-hydraulics of a boiling water reactor",American Nuclear Society, New York, 1993.
- L.S.Tong and J.Weisman,"Thermal analysis of pressurized water reactors",American Nuclear Society, La Grange Park,1996.
- B.Panella,"Reattori nucleari ad acqua leggera", Celid,Torino,1981.

The exam at the end of the course is oral: it regards all the theoretical topics and includes a discussion of the results obtained in the computing laboratory sessions. To be admitted to the exam, the reports of the numerical evaluations must be provided to the teacher at least 5 working days before the exam date. Contributions to the final grade: oral 80%, projects reports 20%.