"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.
Alajo, Ayodeji Babatunde
Castaño, Carlos H.
Nuclear Engineering and Radiation Science
M.S. in Nuclear Engineering
National Academy for Nuclear Training (U.S.)
Missouri University of Science and Technology. Chancellor's Fellowship Program
Missouri University of Science and Technology
x, 58 pages
© 2012 Kelly Christopher Rogers O'Bryant, All rights reserved.
Thesis - Restricted Access
Fuel burnup (Nuclear engineering) -- Measurement
Nuclear reactors -- Safety measures
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Electronic access to the full-text of this document is restricted to Missouri S&T users. Otherwise, request this publication directly from Missouri S&T Library or contact your local library.http://merlin.lib.umsystem.edu:80/record=b9658202~S5
O'Bryant, Kelly Christopher Rogers, "Hot channel determination and burnup analysis of Missouri University of Science and Technology Research Nuclear Reactor" (2012). Masters Theses. 5985.
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