CONTAINMENT PERFORMANCE ANALYSIS OF THE ADVANCED NEUTRON SOURCE REACTOR AT THE OAK-RIDGE-NATIONAL-LABORATORY

This article discusses salient aspects of methodology, assumptions, an d modeling of various features related to the estimation of source ter ms from two conservatively scoped severe-accident scenarios in the Adv anced Neutron Source (ANS) reactor at the Oak Ridge National Laborator y. Various containment configurations are considered for steaming-pool -type accidents and an accident involving molten-core-concrete interac tion. Several design features (such as rupture disks) are examined to study containment response during postulated severe accidents. Also, t hermal-hydraulic response of the containment and radionuclide transpor t and retention in the containment are studied. The results are descri bed as transient variations of source terms for each scenario, which a re to be used for studying off-site radiological consequences and heal th effects for these postulated severe accidents. Also highlighted wil l be a comparison of source terms estimated by two different versions of the MELCOR code.