Methods for calculating dose rates from large nuclear reactor fuel arrays

Recently, progress has been made in calculating dose rates for large arrays of fuel sources using the Monte Carlo neutron photon (MCNP) computer code. The MCNP computer code currently has the capability to model large repeated structures, i.e., the code can duplicate geometry without defining each cell. However, it can be difficult to verify that a fixed source has been accurately modeled. This paper will provide an example of using the repeated structure capability of the MCNP computer code to model a large array of fuel pins and demonstrate how the source locations were verified by graphically displaying the location of the source particles that are generated