Determination of plutonium and uranium content and burnup using six group delayed neutrons

In this study, investigation of spent fuel was performed using six group delayed neutron parameters. Three used fuels (F1, F2, and F11) which are burnt over the years in the core of Missouri University of Science and Technology Reactor (MSTR), were investigated. F16 fresh fuel was used as plutonium free fuel element and compared with irradiated used fuels to develop burnup and Pu discrimination method. The fast fission factor of the MSTR was calculated to be 1.071 which was used for burnup calculations. Burnup values of F2 and F11 fuel elements were estimated to be 1.98 g and 2.7 g, respectively. 239Pu conversion was calculated to be 0.36 g and 0.50 g for F2 and F11 elements, respectively. Keywords: Fuel elements, Burnup, Delay neutrons, Six group parameter