lecture01

lecture01 - 22.05 Reactor Physics Part One Course...

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22.05 Reactor Physics – Part One Course Introduction 1. Instructor: John A. Bernard 2. Organization: Homework (20%) Four Exams (20% each; lowest grade is dropped) Final Exam (3.0 hours) (20%) 3. Text: The text book for this course is: Introduction to Nuclear Engineering, 3 rd Edition, by John Lamarsh. This covers basic reactor physics as part of a complete survey of nuclear engineering. Readings may also be assigned from certain of the books listed below: ± Nuclear Reactor Analysis by A. F. Henry ± Introduction to Nuclear Power by G. Hewitt and J. Collier ± Fundamentals of Nuclear Science and Engineering by J. Shultis and R. Faw ± Atoms, Radiation, and Radiation Protection by J. Turner ± Nuclear Criticality Safety by R. Kneif ± Radiation Detection and Measurement by G. Knoll . Course Objective: To quote the late Professor Allan Henry: 4 “The central problem of reactor physics can be stated quite simply. It is to compute, for any time t, the characteristics of the free-neutron population throughout an extended region of space containing an arbitrary, but known, mixture of materials. Specifically we wish to know the number of neutrons in any infinitesimal volume dV that have kinetic energies between E and E + E and are traveling in directions within an infinitesimal ngle of a fixed direction specified by the unit vector . a If this number is known, we can use the basic data obtained experimentally and theoretically from low-energy neutron physics to predict the rates at which all possible nuclear reactions, including fission, will take place throughout the region. Thus we can 1
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predict how much nuclear power will be generated at any given time at any location in the region.” There are several reasons for needing this information: ± Physical understanding of reactor safety so that both design and operation is done intelligently. ± Design of heat removal systems. ± Fuel management. There are several approaches to the estimation of the neutron population: ± Neutron Life Cycle Analysis: Used for design of the original reactors in the 1950s and early 1960s before computers were available. Very useful for physical understanding. ± One-Velocity Model: A form of diffusion theory in which all neutrons are assumed to have the same speed. Hence, it allows for geometrical and material, but not energy, effects. Useful for designing unreflected (bare) fast reactor cores. ± Diffusion Theory: Design tool for most existing PWRs/BWRs. Also called few-group theory or multi-group theory because the neutrons are treated as in distinct energy ranges or groups. ± Transport Theory: Methods for solving the Boltzmann transport equation (not covered in this course). Monte Carlo Methods: Design technique that is currently Reactor is precisely modeled as to its material and spatial properties. Individual neutron case histories are projected using probability theory. Case histories are run until the statistics are sufficient to assure that an accurate picture of the overall neutron The choice of an ana c 5
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This note was uploaded on 06/02/2009 for the course NSE 22.09 taught by Professor B.forget during the Fall '09 term at MIT.

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lecture01 - 22.05 Reactor Physics Part One Course...

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