Methodology and Determination of Field of View of Neutron and Gamma Detectors in the Atucha Spent Fuel Storage Pool

In this paper we seek to create a model by determining the field of view (FOV) of a detector (i.e. which assemblies contribute to the detector response) in the Atucha-I spent fuel pool. The FOV is determined by solving the adjoint transport equation using the 3- D, parallel PENTRAN (Parallel Environment Neutral-particle TRANsport) Sn code, with the detector cross section as the adjoint source. Reactor criticality was modeled using the MCNP5 (Monte Carlo N-Particle) Monte Carlo code in order to determine the power distribution in each assembly. Depletion calculations were performed for each zone using the ORIGEN-ARP (Automatic Rapid Processing) utility from the SCALE 5.1 (Standardized Computer Analyses for Licensing Evaluation) code package. For the neutron detector, 88% of the response comes from the nearest 4 assemblies (i.e. one assembly in each direction), with 99% from the nearest 16 (i.e. two assemblies in each direction). Results for a uniformly sensitive gamma detector are similar, with 89% of the response coming from the adjacent assemblies. A Monte Carlo calculation using MCNP was performed to benchmark the neutron result, giving a similar result (87% MCNP vs. 88% PENTRAN). Based on these studies, we have developed a database of FOVs as a function of burnup and decay conditions for different detector types, and a methodology/algorithm which uses this database to analyze the response of a detector placed in a spent fuel pool with the aim of detecting gross defects.