Modeling Advanced Neutron Source reactor station blackout accident using RELAP5

Conference · Mon Jan 01 00:00:00 EST 1990

OSTI ID:6978419

Chen, N C.J. ^[1]; Fletcher, C D ^[2]

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1. Oak Ridge National Lab., TN (USA)
2. EG and G Idaho, Inc., Idaho Falls, ID (USA)

The Advanced Neutron Source (ANS) system model using RELAP5 has been developed to perform loss-of-coolant accident (LOCA) and non-LOCA transients as safety-related input for early design considerations. The transients studies include LOCA, station blackout, and reactivity insertion accidents. The small-, medium-, and large-break LOCA results were presented and documented. This paper will focus on the station blackout scenario. The station blackout analyses have concentrated on thermal-hydraulic system response with and without accumulators. Five transient calculations were performed to characterize system performance using various numbers and sizes of accumulators at several key sites. The main findings will be discussed with recommendations for conceptual design considerations. ANS is a state-of-the-art research reactor to be built and operated at high heat flux, high mass flux, and high coolant subcooling. To accommodate these features, three ANS-specific changes were made in the RELAP5 code by adding: the Petukhov heat transfer correlation for single-phase forced convection in the thin coolant channel; the Gambill additive method with the Weatherhead wall superheat for the critical heat flux; and the Griffith drift flux model for the interfacial drag in the slug flow regime. 7 refs., 6 figs., 1 tab.

• Research Organization: Oak Ridge National Lab., TN (USA)
• Sponsoring Organization: DOE/NE
• DOE Contract Number: AC05-84OR21400
• OSTI ID: 6978419
• Report Number(s): CONF-9009219-2; ON: DE90015792; TRN: 90-028876
• Resource Relation: Conference: 1990 joint RELAP5 and TRAC-BWR international user seminar, Chicago, IL (USA), 17-21 Sep 1990
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
99 GENERAL AND MISCELLANEOUS//MATHEMATICS, COMPUTING, AND INFORMATION SCIENCE
LOSS OF COOLANT
R CODES
RESEARCH REACTORS
ACCUMULATORS
BLACKOUTS
COMPUTER CALCULATIONS
HEAT FLUX
HEAT TRANSFER
HYDRAULICS
REACTOR COOLING SYSTEMS
REACTOR SAFETY
T CODES
TRANSIENTS
ACCIDENTS
COMPUTER CODES
CONTAINERS
COOLING SYSTEMS
ENERGY SYSTEMS
ENERGY TRANSFER
FLUID MECHANICS
MECHANICS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
RESEARCH AND TEST REACTORS
SAFETY
TANKS
220900* - Nuclear Reactor Technology- Reactor Safety
220600 - Nuclear Reactor Technology- Research
Test & Experimental Reactors
990200 - Mathematics & Computers