Masters Theses

Abstract

"The development of a computer code was undertaken to calculate the fission product nuclide concentration for spent fuel discharged from a nuclear reactor, as a function of irradiation time, cooling time, neutron flux, and element cross section. The necessary input data was prepared from the basic nuclear data for fission of uranium-235 and plutonium-239. Calculations are also made for total or selective element beta heating and total gamma flux according to user selected input groups. These calculations would be useful in fuel element shielding, cooling, shipping and processing studies"--Abstract, page ii.

Advisor(s)

Edwards, D. R.

Committee Member(s)

Alcorn, Herbert R., 1933-2008
Kim, Hichull

Department(s)

Mining and Nuclear Engineering

Degree Name

M.S. in Nuclear Engineering

Publisher

University of Missouri--Rolla

Publication Date

1968

Pagination

v, 127 pages

Rights

© 1968 Bruce Earl Koopmann, All rights reserved.

Document Type

Thesis - Open Access

File Type

text

Language

English

Library of Congress Subject Headings

Fission products -- Computer programs
Neutron flux -- Measurement
Neutron irradiation -- Measurement

Thesis Number

T 2176

Print OCLC #

6000668

Electronic OCLC #

794078242

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