Reactor Power Distribution Calculation in Research Reactors Using MCNP

In reactor calculations, a detailed 3-D power density distribution is required for core optimization studies and safety analyses. The general Monte Carlo based neutral particle transport tool, MCNP [1], has the capability to obtain detailed power density distribution in a reactor through its criticality calculation mode (KCODE mode). However, there is no standard tally type in MCNP that is able to directly provide total power information (F7 tally only accounts for prompt energy release in fission.). Moreover, because tally results obtained from MCNP are normalized to either fixed source strength (fixed source mode calculation) or total active fission source (keigenvalue mode calculation), some additional efforts are inevitably required to obtain the absolute power factors in the reactor. Power density for a given position in a core is essentially determined by the effective recoverable fission energy deposited in the position. The majority of the fission energy appears as kinetic energy of the fission fragments and is deposited at the point of fission. Over 90% of the recoverable fission energy is deposited directly in the fissile material [2]. In power density calculations with MCNP, we conservatively assume that all the recoverable fission energy is deposited at the point of fission, and the power density is proportional to fission density. Thus the power factors of a position are directly proportional to the fission density at that position. If the power and fuel volume is known, the averaged power density among the fuels can be calculated. Therefore the power density can be obtained by the product of the average power density and the local power factor. With these considerations, the remaining task for power density calculation is to obtain fission density of points under interest.