Simulation of the PKL-G7.1 Experiment in a Westinghouse Nuclear Power Plant Using RELAP5/Mod3.3 (NUREG/IA-0487)

On this page:

• Publication Information
• Abstract

Download complete document

• NUREG/IA-0487 (PDF – 5.50 MB)

Publication Information

Manuscript Completed: January 2019
Date Published: January 2019

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

Universitat Politècnica de València
Camino Vera s/n
46022 Valencia, Spain

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

Availability Notice

Abstract

This paper focuses on the simulation in a Westinghouse design nuclear power plant of the experiment G7.1 conducted at the PKL facility. The test G7.1 consists of a Small Break-Loss of Coolant Accident (SB-LOCA) in the Hot Leg with a total failure of the high-pressure injection system (HPIS). The PKL 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, so some assumptions regarding the scaling and design features of the Westinghouse reactors should be taken into consideration to undertake the simulation.

The postulated additional system failures (no HPIS and no automatic secondary-side cooldown) make necessary to determine the Action Management (AM) measures to prevent core-melt scenario. Under such conditions, the accident mitigation procedure proposed is the manual depressurization of the secondary side steam generators followed by injection from accumulators. The simulation of the transient for a Westinghouse reactor has been performed using RELAP-5 thermal-hydraulic code.