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TRAC L reactor model: Geometry review and benchmarking

Westinghouse Savannah River Co., Aiken, SC (United States), D Griggs, USDOE, Washington, DC (United States), J Cozzuol

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Abstract

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.

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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.

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Available abstract

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.

Key concepts: TRAC, Plenum space, Loss-of-coolant accident, Hydraulics, Nuclear engineering, Accident analysis, Bellows, Engineering

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