Pressurized water reactor steam line break analysis by means of coupled three-dimensional neutronic, three-dimensional core thermohydraulic, and fast running system codes
Journal Article
·
· Nuclear Technology; (United States)
OSTI ID:6963446
- French Atomic Energy Commission, Gif-sur-Yvette (France)
The steam line break (SLB) accident in pressurized water reactors is characterized by a large asymmetric cooling of the core, asymmetric stuck control rods, and large primary coolant flow variations. Because of these space- and time-dependent neutronic and thermal-hydraulic conditions in the core, former SLB analyses that used simplified core models were usually performed with many conservative assumptions. To clarify the complicated behavior of the core, the three-dimensional neutronic code CRONOS-2, the three-dimensional core thermal-hydraulic code FLICA-4, and the system code FLICA-S are completely coupled. The results obtained from the coupled codes indicate that the local thermal-hydraulic feedback effects are important in mitigating neutronic power excursions during SLBs.
- OSTI ID:
- 6963446
- Journal Information:
- Nuclear Technology; (United States), Journal Name: Nuclear Technology; (United States) Vol. 107:1; ISSN 0029-5450; ISSN NUTYBB
- Country of Publication:
- United States
- Language:
- English
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