United States Nuclear Regulatory Commission - Protecting People and the Environment

Assessment of TRACE 5.0 Against ROSA-2 Test 5, Main Steam Line Break with Steam Generator Tube Rupture (NUREG/IA-0505, Revision 1)

On this page:

Download complete document

Publication Information

Manuscript Completed: April 2018
Date Published: April 2019

Prepared by:
S. Gallardo, A. Querol, M. Lorduy, and G. Verdu

Universitat Politecnica de Valencia
Instituto Universitario de Seguridad Industrial, Radiofisica y Medioambiental
Cami de Vera s/n
46022 Valencia, SPAIN

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

Abstract

The purpose of this work is to overview the results obtained by the simulation, using the thermal-hydraulic code TRACE5, of Test 5 (SB-SG-14) in the frame of the OECD/NEA ROSA-2 Project. This test, conducted at the Large Scale Test Facility (LSTF) of the Japan Atomic Energy Agency (JAEA), simulates the thermal hydraulic responses after a PWR Steam Generator Tube Rupture (SGTR) induced by a Main Steam Line Break (MSLB). The result of these simultaneous breaks is a depressurization in both primary and secondary systems because they are connected through the SGTR. The actuation of the Accumulator Injection System was suppressed to keep primary coolant discharge to the Steam Generator secondary-side as low as possible.

The STGR is considered one of the main accidents in nuclear safety due to steam generator reliability and performance are serious concerns in the PWR operation. Through several studies, it has been reported that the severe accident management procedures such as foresee flooding and the primary system depressurization are used to minimize the release from the affected steam generator. These actions may significantly reduce the source term in SGTR accidents

A comparison between experimental and simulation results is provided throughout several graphs, which represent the main thermal-hydraulic variables. In general, TRACE5 shows an acceptable behavior reproducing the experimental data in the entire transient.

Page Last Reviewed/Updated Friday, May 24, 2019