REFERENCES

Foster, Arthur R. and Wright, Robert L. Jr., Basic Nuclear Engineering, 3rd Edition, Allyn and Bacon, Inc., 1977.

Jacobs, A.M., Kline, D.E., and Remick, F. J., Basic Principles of Nuclear Science and Reactors,Van Nostrand Company, Inc., 1960.

Kaplan, Irving, Nuclear Physics,2nd Edition, Addison-Wesley Company, 1962.

Knief, Ronald Allen, Nuclear Energy Technology: Theory and Practice of Commercial Nuclear Power,McGraw-Hill, 1981.

Lamarsh, John R., Introduction to Nuclear Engineering, Addison-Wesley Company, 1977.

Lamarsh, John R., Introduction to Nuclear Reactor Theory,Addison-Wesley Company, 1972.

General Electric Company, Nuclides and Isotopes: Chart of the Nuclides,14th Edition, General Electric Company, 1989.

Academic Program for Nuclear Power Plant Personnel,Volume III, Columbia, MD, General Physics Corporation, Library of Congress Card #A 326517, 1982.

Glasstone, Samuel, Sourcebook on Atomic Energy,Robert F. Krieger Publishing Company, Inc., 1979.

Glasstone, Samuel and Sesonske, Alexander, Nuclear Reactor En ineering,3rd Edition, Van Nostrand Reinhold Company, 1981.

TERMINAL OBJECTIVE

1.0 Without references, EXPLAIN how neutron sources produce neutrons.

ENABLING OBJECTIVES

1.1 DEFINE the following terms:

a. Intrinsic neutron source

b. Installed neutron source

1.2 LIST three examples of reactions that produce neutrons in intrinsic neutron sources.

1.3 LIST three examples of reactions that produce neutrons in installed neutron sources.

TERMINAL OBJECTIVE

2.0 Given the necessary information for calculations, EXPLAIN basic concepts in reactor physics and perform calculations.

ENABLING OBJECTIVES

2.1 DEFINE the following terms:

a. Atom density d. Barn

b. Neutron flux e. Macroscopic cross section

c. Microscopic cross section f. Mean free path

2.2 EXPRESS macroscopic cross section in terms of microscopic cross section.

2.3 DESCRIBE how the absorption cross section of typical nuclides varies with neutron energy at energies below the resonance absorption region.

2.4 DESCRIBE the cause of resonance absorption in terms of nuclear energy levels.

2.5 DESCRIBE the energy dependence of resonance absorption peaks for typical light and heavy nuclei.

2.6 EXPRESS mean free path in terms of macroscopic cross section.

2.7 Given the number densities (or total density and component fractions) and microscopic cross sections of components, CALCULATE the macroscopic cross section for a mixture.

2.8 CALCULATE a macroscopic cross section given a material density, atomic mass, and microscopic cross section.

2.9 EXPLAIN neutron shadowing or self-shielding.

2.10 Given the neutron flux and macroscopic cross section, CALCULATE the reaction rate.

2.11 DESCRIBE the relationship between neutron flux and reactor power.

ENABLING OBJECTIVES (font.)

2.12 DEFINE the following concepts:

a. Thermalization d. Average logarithmic energy decrement

b. Moderator e. Macroscopic slowing down power

c. Moderating ratio

2.13 LIST three desirable characteristics of a moderator.

2.14 Given an average fractional energy loss per collision, CALCULATE the energy loss after a specified number of collisions.

TERMINAL OBJECTIVE

3.0 Without references, EXPLAIN the production process and effects on fission of prompt and delayed neutrons.

ENABLING OBJECTIVES

3.1 STATE the origin of prompt neutrons and delayed neutrons.

3.2 STATE the approximate fraction of neutrons that are born as delayed neutrons from the fission of the following nuclear fuels:

a. Uranium-235 b. Plutonium-239

3.3 EXPLAIN the mechanism for production of delayed neutrons.

3.4 EXPLAIN prompt and delayed neutron generation times.

3.5 Given prompt and delayed neutron generation times and delayed neutron fraction, CALCULATE the average generation time.

3.6 EXPLAIN the effect of delayed neutrons on reactor control.

TERMINAL OBJECTIVE

4.0 Without references, DESCRIBE the neutron energy spectrum for the type of reactor presented in this module.

ENABLING OBJECTIVES

4.1 STATE the average energy at which prompt neutrons are produced.

4.2 DESCRIBE the neutron energy spectrum in the following reactors:

a. Fast reactor

b. Thermal reactor

4.3 EXPLAIN the reason for the particular shape of the fast, intermediate, and slow energy regions of the neutron flux spectrum for a thermal reactor.