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2008 December1 EP6D03 Course Description B. Rouben McMaster University Course EP 6D03 Nuclear Reactor Analysis (Reactor Physics) 2009 Jan.-Apr.

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Presentation on theme: "2008 December1 EP6D03 Course Description B. Rouben McMaster University Course EP 6D03 Nuclear Reactor Analysis (Reactor Physics) 2009 Jan.-Apr."— Presentation transcript:

1 2008 December1 EP6D03 Course Description B. Rouben McMaster University Course EP 6D03 Nuclear Reactor Analysis (Reactor Physics) 2009 Jan.-Apr.

2 2008 December2 Contents Administrative Details Learning Objectives and Outcomes

3 2008 December3 Interaction Information Instructor: Ben Rouben Room: AECL SP1 Conf Room B CRL (Deep River Keys) Videoconf 1) 2) Lectures:Wednesdays 5-8 pm (nominal time) Note: Student active involvement and participation in discussions is very important and strongly encouraged. Questions by Communicate any time. In normal circumstances, I try to respond within hrs or sooner.

4 2008 December4 Study Materials 1) Course notes and postings (see webpage URL next slide) 2) Textbook: Nuclear Reactor Analysis, by James J. Duderstadt & Louis J. Hamilton John Wiley & Sons, Inc. ISBN: ) There are many other Reactor Physics textbooks. 4) You may also find useful material on the CANTEACH website, 5) Note: The material I will present is not necessarily identical to the material in the textbook, and definitely not in the same order. There will be some CANDU-specific material not found in the textbook.

5 2008 December5 Course Webpage The course webpage will be at: I will maintain this webpage with all the latest communications, so I encourage you to visit regularly. To maximise benefit of lectures, read ahead of time, and think about, material to be covered in the next lecture(s).

6 2008 December6 Assignments/Quizzes/Projects Assignments: Assignments with questions/problems, nominally almost every week Project: To be decided in consultation with students. I will assign a special project about half-way in the course, and due by the start of week 12 of the course. The project will/may require FORTRAN programming. Note: Deadlines for Assignments and Project must be met!

7 2008 December7 Grading Tentative: Assignments: 10% Midterm exam: 20% Project: 20% Final exam: 50%

8 2008 December8 Learning Objectives To provide students with knowledge on: Basic concepts and quantities of reactor physics Neutron cycle Neutron transport and diffusion equations, concept of eigenvalue Operator formulation of equations Infinite-lattice concept, solution of 1-group and 2-group diffusion equation in infinite lattice 1-energy-group neutron flux shape in source-sink problems 1-energy-group neutron flux shape in reactors of various geometries 4-factor and 6-factor formulas for reactor multiplication constant Evolution of lattice properties during fuel irradiation Reactivity curve for CANDU lattice Multigroup diffusion theory Neutron kinetics Effect of saturating fission products (Xe-135, Sm-149, etc.) Reactivity devices in CANDU CANDU Fuel Management

9 2008 December9 Learning Outcomes After taking the course, students should be able to: Be familiar with, and use routinely, all the basic concepts of reactor physics, i.e., cross sections, multiplication constants, neutron flux, buckling, irradiation, burnup, eigenvalues, etc. Understand the methods of neutron-transport and neutron-diffusion theory. Be able to solve the 1-group neutron diffusion equation in source-sink problems and in reactors of various geometrical shapes. Understand the isotopic changes in the fuel during fuel irradiation. Be familiar with the lattice reactivity curve for CANDU reactors, and how criticality is maintained over time. Describe and understand the differences between depletion and perturbation calculations. Describe reactivity devices in CANDU reactors and their uses. Understand saturating fission products (e.g., Xe-135, Sm-149) and their effects. Design and code a FORTRAN program to solve the time-independent diffusion equation in a reactor model.

10 2008 December10 Course Outline Introduction to Course Neutron Cycle Operators Neutron Balance Source-Sink Problems Infinite Lattice Finite Reactor in 1 Energy Group Flux Shape in various Reactor Geometries Subcritical Multiplication & Approach to Critical 4-Factor Formula for Reactor Multiplication Constant Neutron Diffusion in 2 Energy Groups contd

11 2008 December11 Course Outline (contd) Finite-Difference Method; Solving the Diffusion Equation Numerically; Course Project Neutron Slowing-Down Kinematics Energy Dependence of Neutron Flux Evolution of Lattice Properties Neutron Fast Kinetics Xe Effects If sufficient time: Lattice Calculations with POWDERPUFS-V CANDU Reactivity Devices CANDU Fuel Management

12 2008 December12 END


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