Abstract
This paper describes work done at the Oak Ridge National Laboratory (ORNL) for evaluating the potential and resulting consequences of a hypothetical criticality accident during refueling of the 330-MW Advanced Neutron Source (ANS) research reactor. The development of an analytical capability is described. Modeling and problem formulation were conducted using concepts of reactor neutronic theory for determining power level escalation, coupled with ORIGEN and MELCOR code simulations for radionuclide buildup and containment transport. Gaussian plume transport modeling was done for determining off-site radiological consequences. Nuances associated with modeling this blast-type scenario are described. Analysis results for ANS containment response under a variety of postulated scenarios and containment failure modes are presented. It is demonstrated that individuals at the reactor site boundary will not receive doses beyond regulatory limits for any of the containment configurations studied.
| Original language | English |
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| Pages | 481-490 |
| Number of pages | 10 |
| State | Published - 1993 |
| Event | Proceedings of the 2nd ASME-JSME Nuclear Engineering Joint Conference. Part 1 (of 2) - San Francisco, CA, USA Duration: Mar 21 1993 → Mar 24 1993 |
Conference
| Conference | Proceedings of the 2nd ASME-JSME Nuclear Engineering Joint Conference. Part 1 (of 2) |
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| City | San Francisco, CA, USA |
| Period | 03/21/93 → 03/24/93 |