Uncertainties in TRAC plenum pressures for the FI phase of a DEGB LOCA

Technical Report · Wed May 01 04:00:00 EDT 1991

DOI:https://doi.org/10.2172/5037132· OSTI ID:5037132

Griggs, D P

The TRAC-PF1/MOD1 code (TRAC) is used to perform best-estimate analyses of certain postulated Design Basis Accidents (DBAs) in SRS production reactors. Currently, the most limiting DBA in terms of reactor power level is an instantaneous double-ended guillotine break (DEGB) loss of coolant accident (LOCA). For this accident, TRAC is used to analyze only the first 5 seconds following the DEGB, which encompasses the Flow Instability (FI) phase of the DBA. The TRAC analysis provides time-dependent plenum and tank bottom pressures for use as boundary conditions in the FLOWTRAN code. The quantification of uncertainty is an important element of determining safe operating power levels for SRS reactors. A detailed methodology for the determination of uncertainty for the FI phase of a DEGB LOCA has been developed. This report presents estimates of the uncertainty in the time-dependent plenum pressures for the DEGB LOCA calculated by TRAC. The plenum pressure uncertainty was estimated by means of comparing TRAC results with steady-state data measured in L Reactor, and confirmed by comparisons with transient LOCA results calculated by an independent group with the RELAP5 code. An overview of the limits methodology is given and discusses the L Reactor data. The methodology for estimating the plenum pressure uncertainty is presented along with the results.

• Research Organization: Westinghouse Savannah River Co., Aiken, SC (United States)
• Sponsoring Organization: DOE; USDOE, Washington, DC (United States)
• DOE Contract Number: AC09-89SR18035
• OSTI ID: 5037132
• Report Number(s): WSRC-TR-90-263; ON: DE92015063
• Country of Publication: United States
• Language: English

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ACCIDENTS
COMPUTER CODES
COOLING SYSTEMS
DATA COVARIANCES
DESIGN BASIS ACCIDENTS
ENERGY TRANSFER
F CODES
FLUID FLOW
FLUID MECHANICS
HEAT TRANSFER
HEAVY WATER MODERATED REACTORS
HYDRAULICS
L REACTOR
LIMITING VALUES
LOSS OF COOLANT
MECHANICS
NATIONAL ORGANIZATIONS
PRODUCTION REACTORS
R CODES
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR SAFETY
REACTORS
SAFETY
SAVANNAH RIVER PLANT
SPECIAL PRODUCTION REACTORS
STABILITY
T CODES
US AEC
US DOE
US ERDA
US ORGANIZATIONS