The course, that the Master level nuclear engineering students follow in the first year, deals with: a) the ability to design the main components and systems of the plant and particularly of the core and primary loops b) the nuclear power plant thermal-hydraulic design computation, with particular reference to light water reactors, taking into account the core design thermal limits, and the mixing phenomena for the fuel elements open channels, c) the nuclear power plant features, with particular reference to innovative and fast breeder reactors, b) the plant decommissioning.

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 the fast breeder reactors as well as the decommissioning options.

The student should have the basic knowledge of the single and two-phase thermalfluidynamics and of nuclear power plants, that can be acquired in the Energetics engineering at Bachelor level and in the course of Thermal and nuclear power plants of the Energetics and Nuclear engineering at Master level.

1. Nuclear reactor design and Regulatory guides; safety and seismic categories; quality assurance. Core thermal hydraulic design methodology depending on the reactor type; interaction between the main sub-projects: neutron, thermal, mechanical, materials project; design of the geometry of the subchannel, of the enrichment and burnup on the basis of neutron, thermal and economic criteria
2. Characteristics of the core of different nuclear reactors, types of subchannels; comparison of the core thermal parameters: power density, heat flux, linear heat power; choice of the parameters with reference to safety margins and evolution of the project, choice of the coolant , enthalpy increase, flow rate, pressure.
3. Thermal core design: fuel rod thermal limits; fuel rod behaviour and failure mechanisms.
4. Nuclear reactor heat generation; thermal power profile for a homogenous not reflected core; effect of control rods, gaps, enrichment, neuton thermal flux. Hot channel factors for heat flux, enthalpy increase and temperature differerence. Hot subchannel analysis. Statistical methods applied to nuclear technology: evaluation of the hot channel factors. 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. Thermalhydraulic design of a PWR core at the beginning and end of life; temperature calculation of the hot channel: coolant, fuel rod wall, cladding, pellet, temperature profiles, conductivity integral, flux depression factor, nucleate boiling on the wall: correlation of Jens-Lottes; Bowring model and other correlations of the THINC code for the region of thermodynamic non-equilibrium; thermal resistance of the gap between pellet and cladding: theory and correlations; evaluation of the thickness of the hot gap; evaluation of the pressure inside the rod at the end of life, and effect of oxide and crud on the temperaturel field; thermal- mechanical design of the rod cladding.
6. PWR thermal 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: W3, Tong factor, grid factor; correlations for low flow and transient conditions.
7. Thermalhydraulic design in boiling water reactors. ISCOR code. Correlations for the flow quality, void fraction evaluation, single and two-phase pressure drops.
8. BWR thermal crisis: dryout mechanisms. Correlations for BWR design: Hench Levy, CISE, GEXL; thermal crisis limited by Flooding (CCFL).
9. Mixing across the open fuel rod subchannels: physical mechanisms, equations of conservation according to the codes COBRA and THINC; parameter of turbulent mixing and correlations; two-phase flow mixing models and correlations of the code MIXER; applications.
10. IAEA International accidents scale. Major nuclear accidents: TMI. 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: EPR, AP1000, PBMR, ESBWR, SMR. Breeder reactors. Four generation nuclear reactors.
12. Decommissioning of nuclear plants. Decontamination and demolition techniques. Waste management.

- Thermalhydralic evaluation of advanced nuclear reactors components.
- PWR core hot channel thermalhydraulic prediction and fuel rod limits verification.

- B. Panella, Appunti.
- C. Lombardi, Impianti nucleari, Città Studi, 2004.
- M. Cumo, Impianti nucleari. Casa Editrice Università 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.

Controlli dell’apprendimento:

Accertamento delle capacità di risoluzione di problemi svolti nel corso.

Modalità di esame:

La valutazione si basa sull'esame orale alla fine del corso e sulle esercitazioni svolte durante l'anno.
Le esercitazioni prevedono l’uso del laboratorio informatico di termoidraulica del Dipartimento di Energetica.