Simulation of the G3.1 Experiment at PKL Facility Using RELAP5/Mod3.3 (NUREG/IA-0486)

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Publication Information

Manuscript Completed: June 2017
Date Published: November 2018

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

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

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Abstract

When a nuclear power plant is in full conditions of operation, or with the primary of reactor coolant system pressurized, one of the most important accidental situation is when a main steam line break occurs, due to that causes a rapid depressurization of the affected steam generator (SG). This depressurization leads to an increment of heat transfer from the primary to the secondary side that cooldown primary side quickly.

This paper focuses on the simulation, using the best estimate code RELAP5/Mod 3.3, of the experiment G3.1 conducted at the PKL facility that simulate a main steam line break. In the experiment G3.1 the physical phenomena to investigate are the assessment of the reactor pressure vessel (RPV) integrity considering PTS (pressurized thermal shock) and the assessment of potential recriticality following entrainment of colder water into the core, thought this work has centered in the thermahydraulic behavior.