Conceptual design loss-of-coolant accident analysis for the Advanced Neutron Source reactor
- Oak Ridge National Lab., TN (United States)
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: (1) 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 of CHF thermal limits at a 95% probability level if the proper mitigation options are provided.
- OSTI ID:
- 5454523
- Journal Information:
- Nuclear Technology; (United States), Journal Name: Nuclear Technology; (United States) Vol. 105:1; ISSN 0029-5450; ISSN NUTYBB
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
22 GENERAL STUDIES OF NUCLEAR REACTORS
220600 -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
COOLING SYSTEMS
DATA
ECCS
ENERGY SYSTEMS
ENGINEERED SAFETY SYSTEMS
HEAVY WATER
HYDROGEN COMPOUNDS
INFORMATION
LOSS OF COOLANT
NUMERICAL DATA
OXYGEN COMPOUNDS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR PROTECTION SYSTEMS
REACTOR SAFETY
REACTORS
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SAFETY
THEORETICAL DATA
WATER