Abstract
This article discusses salient aspects of methodology, assumptions, and modeling of various features related to the estimation of source terms from two conservatively scoped severe-accident scenarios in the Advanced Neutron Source (ANS) reactor at the Oak Ridge National Laboratory. Various containment configurations are considered for steaming-pool-type accidents and an accident involving molten-core-concrete interaction. Several design features (such as rupture disks) are examined to study containment response during postulated severe accidents. Also, thermal-hydraulic response of the containment and radionuclide transport and retention in the containment are studied. The results are described as transient variations of source terms for each scenario, which are to be used for studying off-site radiological consequences and health effects for these postulated severe accidents. Also highlighted will be a comparison of source terms estimated by two different versions of the MELCOR code.
| Original language | English |
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| Pages (from-to) | 205-212 |
| Number of pages | 8 |
| Journal | Nuclear Safety |
| Volume | 35 |
| Issue number | 2 |
| State | Published - Jul 1994 |