Calculation of the neutron flux in the side regions of the prestressed concrete reactor vessel of a high temperature gas cooled reactor

Author

Richard Robert Kent Jr.

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 -- Measurement
Nuclear reactors -- Shielding (Radiation) -- Testing
Gas cooled reactors

Thesis Number

T 2946

Print OCLC #

6022780

Electronic OCLC #

912534901

Link to Catalog Record

http://merlin.lib.umsystem.edu/record=b1066952~S5

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