RESUMEN
Across the globe nuclear utilities are in the process of designing and analysing Independent Spent Fuel Storage Installations (ISFSI) for the purpose of above ground spent-fuel storage primarily to mitigate the filling of spent-fuel pools. Using a conjoining of discrete ordinates transport theory (DORT) and Monte Carlo (MCNP) techniques, an ISFSI was analysed to determine neutron and photon dose rates for a generic overpack, and ISFSI pad configuration and design at distances ranging from 1 to -1700 m from the ISFSI array. The calculated dose rates are used to address the requirements of 10CFR72.104, which provides limits to be enforced for the protection of the public by the NRC in regard to ISFSI facilities. For this overpack, dose rates decrease by three orders of magnitude through the first 200 m moving away from the ISFSI. In addition, the contributions from different source terms changes over distance. It can be observed that although side photons provide the majority of dose rate in this calculation, scattered photons and side neutrons take on more importance as the distance from the ISFSI is increased.
Asunto(s)
Modelos Estadísticos , Reactores Nucleares/instrumentación , Monitoreo de Radiación/métodos , Protección Radiológica/instrumentación , Residuos Radiactivos/análisis , Administración de Residuos/instrumentación , Simulación por Computador , Análisis de Falla de Equipo , Método de Montecarlo , Neutrones , Fotones , Dosis de Radiación , Monitoreo de Radiación/normas , Protección Radiológica/métodos , Dispersión de Radiación , Administración de Residuos/métodosRESUMEN
The inventory of radioactivity that must be considered in the decommissioning of a typical 1000 MWe Spanish pressurised water reactor (PWR) was investigated as part of a generic plant decommissioning study. Analyses based on DORT models (in both R-Z and R-theta geometries) were used with representative plant operating history and core power distribution data in defining the expected neutron environment in regions near the reactor core. The activation analyses were performed by multiplying the DORT scalar fluxes by energy-dependent reaction cross sections (based on ENDF/B-VI data) to generate reaction rates on a per atom basis. The results from the ORIGEN2 computer code were also used for determining the activities associated with certain nuclides where multi-group cross section data were not available. In addition to the bulk material activation of equipment and structures near the reactor, the activated corrosion-product (or 'crud') deposits on system and equipment surfaces were considered. The projected activities associated with these sources were primarily based on plant data and experience from operating PWR plants.