The course deals with fluid flow phenomena in energy exchanges with emphasis on boiling two-phase flows.

It starts with fundamental concepts of nuclear reactor safety, in brief reference to the Fukushima Daiichi Nuclear Plant accident of 2011, where the importance of thermal hydraulics phenomena is emphasized. The course shall focus mainly on Light Water-cooled Reactors (LWRs - PWRs and BWRs) and Sodium-cooled Fast Reactors (SFRs) under normal and accident conditions.

After a cursory look at the nuclear reactor core thermal hydraulics (TH) design methods of pressure drop and heat transfer modeling, verification/validation and uncertainty quantification, we delve into a detail of the two-phase flow representation and its associated physical modeling. Emphasis will be placed on the pool boiling and convective boiling heat transfer mechanisms, condensation, critical heat flux (CHF), and post-CHF heat transfers. Boiling flow natural circulation (NC) in a closed circuit for Boiling Water Reactors and low heat flux sodium boiling NC phenomena will be understood through exercises. The method is applicable for any types of Gen-IV reactors including Molten Salt Reactors. One of the advanced and most robust TH calculation methods for nuclear fuel rod bundles is exemplified by the subchannel analysis and will be described in completing the course.

During the course, a number of exercises will be performed and explained in the class and homeworks will be assigned to the students.

The students will know and comprehend:  

• a wide spectrum of advanced thermal hydraulics and associated nuclear reactor safety;
• modern approaches to the design of advanced nuclear power from the calculation of average values to local, which is particularly important when designing intense nodes of the reactor core, steam generators, heat exchangers.

The students will able to: 

• elaborate detailed knowledge of distributed parameters, i.e., detailed coolant flow and temperature distributions, the surface temperature distribution and uncertainties to consider the account for possible deviations from the calculated actual values. Such an idea is constantly being implemented and put into practice in the current engineering practices;
• make their own judgement on how safe the nuclear reactors of their concern;
• discuss and communicate with good property of language and scientific terminology.

Nuclear Reactor Safety: Start with a cursory look at the general principle of nuclear reactor safety, the defence-in-depth principle and its practice in nuclear power plant safety, concepts of PSA, we learn the lessons from the Fukushima Daiichi Nuclear Station Accident that took place March 11, 2011. The role of thermal hydraulics is highlighted focusing on LOCA, a typical design basis accident (DBA) of LWR.

Nuclear Reactor Thermal Hydraulics Fundamentals and Exercises: With knowledge assumed on single-phase flow transport equations and heat transfer mechanisms of conduction, convection and radiation heat transfers: 1) Fundamental parameters to describe boiling and two-phase flows; 2) Heat balance and enthalpy (temperature) distributions; exercises in non-boiling (PWR) and boiling (BWR) channel exit quality of nuclear fuel rod assemblies; 3) Verification and validation, uncertainty quantification and propagation in thermal hydraulics evaluation; 4) Hot spot factors and subfactors.

Two-Phase Flow Modeling: Conservation equations described by the two-fluid model, homogeneous equilibrium mixture model, and the drift flux model. Pressure drops with two-phase flow multiplier concepts for HEM and two-fluid model description. Relationship between void fraction and steam quality is derived from the drift flux model.

Pool Boiling Phenomena: Discussions are given to heat transfers along the boiling curves in pool boiling on bubble nucleation, onset of nucleate boiling on a heated surface, bubble growth and departure, nucleate boiling, critical heat flux (CHF) phenomena and film boiling and minimum film boiling temperature.

Condensation: Physics of the process and heat transfer; Drop-wise and film-wise condensation; Condensation of vapor/non-condensable gas mixture; Direct contact condensation.

Convective Boiling, Critical Heat Flux and Post CHF Heat Transfers: Focus on the method to determine the wall temperature or heat flux under various convective boiling heat transfer modes illustrated by boiling curves and regions of heat transfer including single-phase liquid phase, onset, subcooled and developed nucleate boiling regions, and saturated boiling and two-phase forced convection regions; Burnout and critical heat flux (CHF) phenomena – models for DNB and film dryout; Design margin to the critical condition; Post CHF heat transfer in inverted and dispersed annular flow regimes.

Applications of Two-Phase Flow Models: Exercises on boiling two-phase flow natural circulation in a closed circuit for BWRs under operating conditions and on low heat flux sodium boiling natural circulation and flow excursion CHF with the Ledinegg instabilities; Subchannel analysis method for nuclear fuel rod bundles, will be described in detail in concluding the ATHANS course.

Thermal hydraulics is one of the most important constituents in extremely complex nuclear reactor system design and safety. In this regard, students are assumed to have fundamental knowledge on both reactor physics and thermo-hydraulics as exemplified in the course “DYNAMICS AND CONTROL OF NUCLEAR PLANTS” by Prof. A. Cammi in the 2017/2018 2^nd Semester. 

The evaluation consists of an oral examination being aimed at verifying the students’ acquisition and understanding of the course topics and the key concepts explained during the class hours.  

The examination will also focus on discussing the assigned homework to test the ability to describe and communicate in a professional manner the theoretical fundamentals of nuclear reactor thermal hydraulics and results of the assignment. 

Chapters 11 to 14 are relevant to boiling two-phase flow heat transfers and used in the lecture.

Chapter 7 Heat transfer in nuclear thermal hydraulics, Kirillov, P. and Ninokata, H.