Numerical approach for predicting the decay heat contribution of curium isotopes in the mixed oxide nuclear fuel

The mixed oxide nuclear fuel (MOX) of U and Pu contains several percent of fission products and minor actinides, such as neptunium, americium and curium. It is important to determine accurately the decay heat from curium isotopes as they contribute significantly in the MOX fuel. This amount of heat will be always weighted by the content of Cm of the fuel. The HEATKAU V.I has been successfully used before to estimate the decay heat of 235U and 239Pu fuels in case of thermal and fast neutron induced fission. The validity of this code has been verified via comparisons against well established calculation techniques such as hybrid and summation methods. The calculations show how accurate and fast is this code than well known codes such as ALEPH and OREGIN 2.2 as mentioned in our previous publication (Nafee et al., 2013). Therefore, in the present paper, we used our HEATKAU V.I inventory code to predict the decay heat from curium isotopes. The calculation procedure of this code is based on the numerical solution of coupled linear differential equations of fissioning systems that describe decays and buildups of many nuclides to calculate the decay heat produced after shutdown.