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EP6D03 Course Description

EP6D03 Course Description. B. Rouben McMaster University Course EP 6D03 Nuclear Reactor Analysis (Reactor Physics) 2009 Jan.-Apr. Contents. Administrative Details Learning Objectives and Outcomes. Interaction Information. Instructor: Ben Rouben Room: AECL SP1 Conf Room B

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EP6D03 Course Description

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

  2. Contents • Administrative Details • Learning Objectives and Outcomes

  3. Interaction Information • Instructor: Ben Rouben • Room:AECL SP1 Conf Room B CRL (Deep River Keys) Videoconf • e-mail: 1) roubenb@mcmaster.ca 2) roubenb@alum.mit.edu • Lectures: Wednesdays 5-8 pm (nominal time) • Note: Student active involvement and participation in discussions is very important and strongly encouraged. • Questions by e-mail: Communicate any time. In normal circumstances, I try to respond within 24-48 hrs or sooner.

  4. 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: 0-471-22363-8 3) There are many other Reactor Physics textbooks. 4) You may also find useful material on the CANTEACH website, http://candu.canteach.org 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. Course Webpage • The course webpage will be at: http://www.nuceng.ca/br_space/2009-01_6d03/6d03_home.htm • 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. 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. Grading Tentative: • Assignments: 10% • Midterm exam: 20% • Project: 20% • Final exam: 50%

  8. 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. 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. 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 cont’d

  11. Course Outline (cont’d) • 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. END

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