Record Details
Probabilistic risk assessment of a vented fuel system utilized in a generation IV gas-cooled fast reactor nuclear power plant
ScholarsArchive at Oregon State University
| Field | Value |
|---|---|
| Title | Probabilistic risk assessment of a vented fuel system utilized in a generation IV gas-cooled fast reactor nuclear power plant |
| Names |
Lenhof, Renae
(creator) Klein, Andrew C. (advisor) |
| Date Issued | 2014-09-12 (iso8601) |
| Note | Graduation date: 2015 |
| Abstract | A limited scope Probabilistic Risk Assessment (PRA) of a vented nuclear fuel system for a Generation IV Gas-Cooled Fast Reactor Plant was performed. The goal of the study was to better understand the safety and licensing implications of vented fuel technology. A Level 1 PRA was performed to determine the possible and probable failure mechanisms of the vented fuel fission product collection and auxiliary systems. A Level 2 PRA was performed to determine the possible releases from the vented fuel collection system under normal and abnormal operating conditions. Finally, a Level 3 PRA was performed to understand the consequence of a release of fission products from the vented fuel collection system to the population and environment. This work has found that the most probable system damage states are a release from the vented fuel system within the containment building and within the auxiliary building. As a result of the reduced fission product boundaries present in the event of a release within the auxiliary building, consequences of these events prove to be more severe. Three source term groups were found to have exceedingly higher dose consequences with respect to their estimated frequencies. As a result, the present analysis indicates that additional design and analysis on the vented fuel system is required. It should be noted that the present work operates under several highly conservative assumptions. Consequently, the vented fuel system, as modeled, may be safely operated without design changes pending the refinement of the models used to determine source term estimates and atmospheric transport. Specifically, the conservative modeling assumptions that have the potential to greatly reduce the estimated consequences, if refined, include the fraction of the activity within the adsorber beds that is released under accident conditions, the specific attenuation factors applied to the various fission product boundaries, and the atmospheric conditions selected for the atmospheric transport model. In addition, design changes such as an increase in the Exclusion Area Boundary (EAB) or placement of the entire vented fuel system within the containment building may further reduce the consequences of a release. This work was performed using funding received from the DOE Office of Nuclear Energy’s Nuclear Energy University Program. |
| Genre | Thesis/Dissertation |
| Access Condition | http://creativecommons.org/licenses/by-nd/3.0/us/ |
| Topic | PRA |
| Identifier | http://hdl.handle.net/1957/52593 |
