RELAP5 and TRACE Calculations of LOCA in PWR (NUREG/IA-0479)

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Publication Information

Manuscript Completed: November 2016
Date Published: June 2017

Prepared by:
Andrej Prošek

Jožef Stefan Institute
Jamova cesta 39
Sl-1000 Ljubljana,

Kirk Tien, NRC Project Manager

Division of Systems Analysis
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

Prepared as part of:
The Agreement on Research Participation and Technical Exchange
Under the Thermal-Hydraulic Code Applications and Maintenance Program (CAMP)

Published by:
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001


Availability Notice

The accident at the Fukushima Dai-ichi nuclear power plant in 2011 demonstrated that external events could cause loss of all safety systems. In the Europe stress tests were performed and the need was identified to further improve the safety of the existing operating reactors. Therefore the safety upgrade programs were started. The objective of this study was to demonstrate that developed input model of two-loop pressurized water reactor (PWR) for TRACE thermal-hydraulic systems code can be used for independent calculations to be compared with RELAP5 computer code calculations. For demonstration the response of PWR to loss-of-coolant accident (LOCA) break spectrum from 10.16 cm (4 inch) to 30.48 cm (12 inch) was simulated. Only passive accumulators were assumed available. For calculations the latest TRACE Version 5.0 Patch 4 and RELAP5/MOD3.3 Patch 4 using both break flow models were used. The results showed that RELAP5 calculations using different break flow models are rather similar, therefore also other parameters are similar. The accumulators discharge was faster in TRACE calculation than in RELAP5 calculations. It can be concluded that different accumulator discharge influencing the break flow seems to be the largest contributor to the differences in the results between RELAP5 and TRACE.

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