Open Access. Powered by Scholars. Published by Universities.®
<p>Fuel burnup (Nuclear engineering) -- Measurement<br />Nuclear reactors -- Safety measures<br />Nuclear fuels</p>
Articles 1 - 1 of 1
Full-Text Articles in Engineering
Hot Channel Determination And Burnup Analysis Of Missouri University Of Science And Technology Research Nuclear Reactor, Kelly Christopher Rogers O'Bryant
Hot Channel Determination And Burnup Analysis Of Missouri University Of Science And Technology Research Nuclear Reactor, Kelly Christopher Rogers O'Bryant
Masters Theses
"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 …