TRAC L reactor model: Geometry review and benchmarking

Technical Report · Wed Aug 01 00:00:00 EDT 1990

DOI:https://doi.org/10.2172/10158702· OSTI ID:10158702

Griggs, D P ^[1]; Cozzuol, J M ^[2]

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1. Westinghouse Savannah River Co., Aiken, SC (United States)
2. Idaho National Engineering Lab., Idaho Falls, ID (United States)

The analysis of the Design Basis Loss of Coolant Acident (LOCA) for Savannah River Site (SRS) reactors involves the best estimate reactor system thermal-hydraulics code TRAC-PFI/MOD1. Power levels for the L-3.1 and P-10.2 subcycles were determined based, in part, on TRAC analyses of the first few seconds of a plenum inlet break LOCA. The TRAC code is currently being used to analyze reactor system response for the Double Ended Guillotine Break (DEGB) LOCA, the Expansion Joint Bellows Break LOCA, the Loss of Pumping Accident (LOPA), and the Pump Shaft Break event. Currently, the DEGB LOCA analysis is performed with TRAC only for the flow instability (FI) phase of the accident. This analysis provides input to the determination of operating power limits for the K-14.1 subcycle.

• Research Organization: Savannah River Site (SRS), Aiken, SC (United States)
• Sponsoring Organization: USDOE, Washington, DC (United States)
• DOE Contract Number: AC09-89SR18035
• OSTI ID: 10158702
• Report Number(s): WSRC-TR-90-32; ON: DE92016494
• Resource Relation: Other Information: PBD: Aug 1990
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
99 GENERAL AND MISCELLANEOUS//MATHEMATICS, COMPUTING, AND INFORMATION SCIENCE
PRODUCTION REACTORS
LOSS OF COOLANT
FLUID FLOW
STABILITY
SAVANNAH RIVER PLANT
T CODES
HEAT TRANSFER
HYDRAULICS
REACTOR SAFETY
REACTOR COOLING SYSTEMS
220900
220600
990200
RESEARCH
TEST
TRAINING
PRODUCTION
IRRADIATION
MATERIALS TESTING REACTORS
MATHEMATICS AND COMPUTERS