United States Nuclear Regulatory Commission - Protecting People and the Environment

Simulation of the LSTF-PKL Counterpart G7.1 Test at PKL Facility Using TRACE 5 (NUREG/IA-0488)

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Publication Information

Manuscript Completed: January 2019
Date Published: January 2019

Prepared by:
F. Sánchez, S. Carlos, J. F. Villanueva, and S. Martorell

Universitat Politècnica de València
Camino 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 PKL-2 test program is investigating safety issues relevant for current pressurized water reactor (PWR) plants as well as for new PWR design concepts and will focus on complex heat transfer mechanisms in the steam generators and boron precipitation processes under postulated accident situations.

These issues are being investigated by means of thermal-hydraulic experiments that are conducted at the Primärkreislauf-Versuchsanlage (primary coolant loop test facility) PKL. This facility is owned and operated by AREVA NP and is situated in Erlangen, Germany.

AREVA NP has, for a number of years, conducted valuable experiments on reactor thermal-hydraulics in the PKL facility, including earlier experiments carried out in the framework of the SETH project (2001-2003) and of the PKL-1 Project (2004-2007).

The first category includes tests addressing the heat transfer mechanisms in the steam generators in the presence of nitrogen, steam and water, in both vertical and horizontal steam generators. Cooldown procedures in the case where steam generators have partly dried out on the secondary side have also been covered. A further topic addresses the heat transfer in the steam generators under reflux condenser conditions (e.g. fast secondary side depressurization). Fast cooldown transients (with water filled reactor coolant system) such as main steam line break, completed by tests on mixing of hot and cold water in the RPV downcomer and the lower plenum are also considered in the test program. Further investigation address boron precipitation processes in the core following large break loss of coolant accidents. G7: Counterpart Test with ROSA / LSTF on small break LOCA with Accident Management procedures (one test performed in July 2011).

This paper focuses on the simulation, using the best estimate code TRACE, of the experiment G7.1 conducted at the PKL facility. The OECD-PKL2 test G7.1 is a hot leg Small Break-Loss of Coolant Accident (SB-LOCA) with a total failure of the High-Pressure Safety Injection (HPSI). The PKL test facility simulates a KWU 1300 MWe pressurized water reactor with all elevations scaled 1:1 and with volume and power scaled by a factor of 1:145. This test had been previously performed in ROSA/LSTF facility in order to compare results between facilities and to analyze different plant configurations and scaling effect.

The postulated additional system failures (no HPSI, no automatic secondary-side cooldown) return a course of events that necessitates Action Management (AM) measures to prevent core-melt scenario. A fast secondary-side depressurization initiated after occurrence of core uncovery was employed, as AM measure for restoration of the secondary side heat sink aiming for a fast reduction of the primary pressure. The reduction of the primary pressure down to ACC injection pressure then effectuates the transition to the low-pressure phase with the Low-Pressure Safety Injection (LPSI) active.

Page Last Reviewed/Updated Tuesday, January 22, 2019