Masters Theses

Abstract

"A burnup analysis has been performed on the Missouri University of Science and Technology (Missouri S&T) Research Nuclear Reactor (MSTR). With use of the Monte Carlo neutronics depletion code MCNPX, burned material was input into a neutronics model (burned model) in order better simulate MSTR core characteristics. Simulated burnup values of ²³⁵U for the past twenty years of MSTR operations totaled 14.266 grams, slightly less than the 14.527 grams reported by the reactor staff. A distribution of ²³⁵U was simulated and a burnup map was developed.

Using the updated fuel material, the hot channel of the current configuration of the MSTR was determined by tallying energy deposition throughout the core. The hot channel is the limiting design factor of the core and is important in reactor safety analysis. The hot channel factor (ratio of hottest to average fuel plate) was determined to be 1.71 for the Core Configuration 120. Relative axial flux counts were simulated and experimentally measured. Simulated values were calculated with a maximum error of 8.54%compared to measured values. The burned model was tested to determine its improvement over the old model (clean model). In determining the effective multiplication factor at experimentally measured critical rod heights of the MSTR at 10 W, the burned model demonstrated a minimum 38% improvement over the clean model. The burned model was minimally 35% closer to simulating the experimentally measured shutdown margin and 46% closer to the excess reactivity. Incorporating burned material provided a more accurate model of the MSTR which can be in future core analysis"--Abstract, page iii.

Advisor(s)

Alajo, Ayodeji Babatunde

Committee Member(s)

Usman, Shoaib
Castaño, Carlos H.

Department(s)

Nuclear Engineering and Radiation Science

Degree Name

M.S. in Nuclear Engineering

Sponsor(s)

National Academy for Nuclear Training (U.S.)
Missouri University of Science and Technology. Chancellor's Fellowship Program

Publisher

Missouri University of Science and Technology

Publication Date

Fall 2012

Pagination

x, 58 pages

Note about bibliography

Includes bibliographical references (pages 56-57).

Rights

© 2012 Kelly Christopher Rogers O'Bryant, All rights reserved.

Document Type

Thesis - Open Access

File Type

text

Language

English

Subject Headings

Fuel burnup (Nuclear engineering) -- MeasurementNuclear reactors -- Safety measuresNuclear fuels

Thesis Number

T 10125

Print OCLC #

842963487

Electronic OCLC #

908944296

Download

Share

 
COinS