Masters Theses

Abstract

"A thermal hydraulic analysis of a typical hot channel of a Dresden 2 class boiling water reactor is studied for possible severe local overheating. The COBRA-II thermal hydraulic analysis code is modified to include critical heat flux calculations for each subchannel. The effects of flow distribution, inlet mass velocity variations, dimensional tolerances, enrichment variations, and input parameters are examined in detail. Bulk channel results are in good agreement with the published data, but the assembly wall side of the corner fuel rod has a minimum critical heat flux ratio of less than unity for a number of the situations examined. In some cases the critical heat flux ratio is less than one for a 2 to 4 foot segment of the rod. In these cases, conditions have changed from nucleate to film boiling. The conclusions of this study is that fuel rod damage may occur due to severe local overheating as a result of inadequate cooling. This is particularly true for the inside corner rod in the region from 72 to 84 inches up the rod"--Abstract, page ii.

Advisor(s)

Edwards, D. R.

Committee Member(s)

Bolon, Albert E., 1939-2006
Culp, Archie W., Jr.

Department(s)

Nuclear Engineering and Radiation Science

Degree Name

M.S. in Nuclear Engineering

Publisher

University of Missouri--Rolla

Publication Date

1974

Pagination

vii, 76 pages

Note about bibliography

Includes bibliographical references (pages 57-58).

Rights

© 1974 Donald Lee Moffett, All rights reserved.

Document Type

Thesis - Open Access

File Type

text

Language

English

Subject Headings

Boiling water reactors
Nuclear fuel rods -- Testing
Heat flux -- Research

Thesis Number

T 2947

Print OCLC #

6022818

Electronic OCLC #

913834703

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