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- W151917520 abstract "A RELAP5 system model for the Advanced Neutron Source Reactor has been developed for performing conceptual safety analysis report calculations. To better represent thermal-hydraulic behavior of the core, three specific changes in the RELAP5 computer code were implemented: a turbulent forced-convection heat transfer correlation, a critical heat flux (CHF) correlation, and an interfacial drag correlation. The model consists of the core region, the heat exchanger loop region, and the pressurizing/letdown system region. Results for three loss-of-coolant accident analyses are presented: (a) an instantaneous double-ended guillotine (DEG) core outlet break with a cavitating venturi installed downstream of the core, (b) a core pressure boundary tube outer wall rupture, and (c) a DEG core inlet break with a finite break-formation time. The results show that the core can survive without exceeding the flow excursion or CHF thermal limits at a 95% probability level if the proper mitigation options are provided." @default.
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- W151917520 date "1994-01-01" @default.
- W151917520 modified "2023-09-23" @default.
- W151917520 title "Conceptual Design Loss-of-Coolant Accident Analysis for the Advanced Neutron Source Reactor" @default.
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- W151917520 doi "https://doi.org/10.13182/nt94-a34914" @default.
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