Deterministic methods for neutron transport - 5PMGSND0
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Number of hours
- Lectures : 8.0
- Tutorials : 8.0
- Laboratory works : 0
- Projects : 0
- Internship : 0
- Written tests : 0
ECTS : 1.5
Goals
Introduction to one numerical method used in deterministic neutron transport codes : collision probabilities.
Understanding the industrial neutron calculation scheme.
Getting experience with the practical use of a deterministic neutron transport code for simplified reactor physics calculations.
Calculation of temperature coefficients, homogeneisation, fuel burn up calculations.
Contact Adrien BIDAUD
Content Neutron Transport theory :
From transport to diffusion equation : assumptions and approximations.
Example of a transport solution method : Collision Probabilities.
Preparing diffusion cross section : energy condensation, flux-volume homogeneisation, transport-diffusion equivalence.
Fuel burn-up calculation.
Fuel management
Industrial calculation scheme
Prerequisites- Reactor physicis : criticality, temperature feedbacks, fuel management
- Neutronics : slowing-down equation, resonance self-shielding, transport equation, diffusion model
- Numerical methods : PDEs, numerical quadrature methods, solving numerically linear systems.
Tests TD = Exercice session evaluation (3*10%)
projet = Simulation project : Fuel burn-up of simple fuel cells (70%)
Contrôle continu : 3*10%
Projet = 70%
Additional Information This course brings 2.0 ECTS to students in TU Simulations
This course brings 1.0 ECTS to students in UE Simulations JUAS
Curriculum->Double-Diploma Engineer/Master->Semester 9
Curriculum->Engineering degree->Semester 9
Bibliography DRAGON User Manual : TECHNICAL REPORT IGE–174 Rev. 12 (Release 3.06L)
"La Neutronique", Monographies CEA/DEN, 2013, ISBN 978-2-281-11371-6
Applied Reactor Physics, Alain Hebert, Presses internationales Polytechnique, 2009
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Date of update March 13, 2019