United States Nuclear Regulatory Commission - Protecting People and the Environment

Feedwater Line Break Analysis Using RELAP5/MOD3.3 for Steam Generator Blowdown Load Assessment (NUREG/IA-0473)

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Publication Information

Manuscript Completed: February 2016
Date Published: November 2016

Prepared by:
Han-Sik JUNG, Du-Ho HONG, Jong-In KIM, Ae-Ju CHEONG*, Kyung-Won LEE*

Doosan Heavy Industry and Construction
22, Doosan Volvo-ro, Seongsan-gu,
Changwon, Gyeongnam, 642-792, Korea

*Korea Institute of Nuclear Safety
62 Gwahak-ro, Yuseong-gu
Daejeon, 34142, Korea

K. 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:
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

Availability Notice


The RELAP5/MOD3.3 is generally used for best-estimate transient simulation of light water reactor coolant systems during postulated accidents in the Light Water Reactor (LWR). The RELAP5/MOD3.3 code is based on a non-homogeneous, and non-equilibrium model for a two phase system that is solved by a fast, partially implicit numerical scheme to permit economical calculation of system transients. This code is suitable for the analysis of transients and postulated accidents in LWR systems, including both large- and small-break loss of coolant accidents, as well as for the full range of operational transients.

For the evaluation of structural integrity for the steam generator in the Pressurized Water Reactor (PWR), the postulated accidents, such as the Feedwater Line Break (FLB) in the Advanced Power Reactor (APR1400) at the Korean domestic plants, are considered Design Basis Events (DBE). In order to evaluate the structural integrity of a steam generator during the FLB, the data for the thermo-hydraulic velocity, density and pressure are needed.

This study was performed to calculate thermal hydraulic parameters, such as thermo-hydraulic velocity, density and pressure, using the RELAP5/MOD3.3 code for the structural evaluation of the steam generator internals during the postulated FLB accidents.

The calculation results were verified by comparing with experimental data generated from the experimental facility ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation).

Page Last Reviewed/Updated Tuesday, November 08, 2016