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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

Curriculum->GEN->Semester 9
Curriculum->Double Diploma GEN-EN->Semester 9
Curriculum->Double-Diploma Engineer/Master->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

Université Grenoble Alpes