Masters Theses
Abstract
"Neutron flux in the side regions of the High Temperature Gas Cooled Reactor's (HTGR) Prestressed Concrete Reactor Vessel (PCRV) was determined using the Monte Carlo method. One quarter of the PCRV was modeled, with all major structures inside it, e.g. steam generator, auxiliary cooling loop, etc. imaged at their exact position and size. Initial neutron energies range from .001 eV to 3 MeV. Three neutron interactions were considered: absorbtion, elastic scattering, and inelastic scattering.
The flux was calculated for two different PCRV temperatures, 155° F and 647° F. The thermal flux showed a minor increase at the higher temperature. In general, the results agree with previous calculations which were made using the transport method"--Abstract, page ii.
Advisor(s)
Tsoulfanidis, Nicholas
Committee Member(s)
Edwards, D. R.
Malisch, Ward R.
Department(s)
Nuclear Engineering and Radiation Science
Degree Name
M.S. in Nuclear Engineering
Publisher
University of Missouri--Rolla
Publication Date
1974
Pagination
ix, 75 pages
Note about bibliography
Includes bibliographical references (pages 50-51).
Rights
© 1974 Richard Robert Kent, Jr., All rights reserved.
Document Type
Thesis - Open Access
File Type
text
Language
English
Subject Headings
Neutron flux -- MeasurementNuclear reactors -- Shielding (Radiation) -- TestingGas cooled reactors
Thesis Number
T 2946
Print OCLC #
6022780
Electronic OCLC #
912534901
Recommended Citation
Kent, Richard Robert Jr., "Calculation of the neutron flux in the side regions of the prestressed concrete reactor vessel of a high temperature gas cooled reactor" (1974). Masters Theses. 3455.
https://scholarsmine.mst.edu/masters_theses/3455