PHYSICS AND NUMERICAL MODELS OF NUCLEAR REACTORS
Academic year2016/17
CourseNUCLEAR ENGINEERING
Code518II
Credits12
PeriodSemester 1 & 2
LanguageEnglish
| Modules | Area | Type | Hours | Teacher(s) |
| FISICA DEI REATTORI NUCLEARI | ING-IND/18 | LEZIONI | 60 | |
| MODELLI NUMERICI PER REATTORI NUCLEARI | ING-IND/18 | LEZIONI | 60 | |
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Knowledge
The student who successfully completes the course will have the ability to understand the neutron transport phenomena typical of a nuclear reactor core, with reference both to the static and dynamic effects important for the reactor core design. He will be able to demonstrate a solid knowledge of the neutron diffusion and transport theories, particularly as it concerns their application to engineering problems like the determination of the multiplication constant of a reactor core or the calculation of a reactor fuel cell. He will be also aware of the mathematical and numerical tools that are at the basis of the typical reactor core calculations.
Assessment criteria of knowledge
During the oral examination the student will be assessed on his/her demonstrated ability to discuss with rigor the main course contents using the appropriate terminology. It is also expected that he/she will show a good ability in making connection between the different topics of the course.
Methods:
Teaching methods
Delivery: face to face
Learning activities:
- attending lectures
- participation in seminar
- participation in discussions
- Laboratory work
Attendance: Advised
Teaching methods:
- Lectures
- Seminar
- laboratory
Syllabus
The course covers the following aspects: the continuity equation, the Fick’s law and the diffusion equation of the neutrons; steady-state neutron diffusion problems with fixed source; the Green function of the diffusion equation; the neutron slowing down and the resonance absorption; the multiplication constant of a critical reactor with one or more energy groups; the delayed neutrons and the kinetic of a homogeneous reactor core; typical strategies to solve the criticality problems with one or more energy groups; solution of linear algebraic equation systems with direct or iterative methods (Jacobi, Gauss-Seidel, SOR, ADI,...); numerical solution of multigroup neutron kinetics problems with delayed neutrons; the integro-differential transport equation and its spherical harmonics approximation; the integral transport equation and its derivation from the integro-differential form; the collision probability method; the discrete ordinate method; the ray effect.
Bibliography
Teaching materials provided by the teacher.
Recommended reading includes:
J.R. Lamarsh, Nuclear Reactor Theory, Addison Wesley Publishing;
E.E. Lewis, W.F. Miller, Computational Methods of Neutron Transport, Wiley-Interscience Publication;
G.I. Bell, S. Glasstone, Nuclear Reactor Theory, Van Nostrand Reinhold Company;
A. Hebert; Applied Reactor Physics, Presses Internationales Polytechnique.
Updated: 14/11/2016 17:27
