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APEX-AP1000 Confirmatory Testing To Support AP1000 Design Certification (Non-Proprietary) (NUREG-1826)

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Publication Information
Abstract

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NUREG-1826 (PDF - 2.46 MB)

Publication Information

Manuscript Completed: June 2005
Date Published: August 2005

Prepared by
K.B. Welter and S.M. Bajorek
Advanced Reactors and Regulatory Effectiveness Branch
Division of Systems Analysis and Regulatory Effectiveness
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

José Reyes, Jr., Brian Woods, John Groome, John Hopson,
Eric Young, John DeNoma, and Kent Abel
Department of Nuclear Engineering
Institute of Thermal-Hydraulics
Oregon State University
Corvallis, OR 97601

Gene S. Rhee, NRC Project Manager

Prepared for
Division of Systems Analysis and Regulatory Effectiveness
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

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Abstract

This report summarizes the confirmatory testing performed using the Advanced Plant Experiment (APEX) facility at Oregon State University (OSU). The APEX is a unique thermal-hydraulic integral system test facility, which is used to assess the performance of passive safety systems and was modified to accurately represent the design of the Westinghouse AP1000 advanced passive nuclear reactor. The U.S. Nuclear Regulatory Commission (NRC) sponsored eight beyond-design-basis accident (DBA) tests in the APEX facility, which were successfully completed from June 2003 through July 2004 and are discussed in detail in this report. Those eight beyond-DBA tests investigated scenarios with two or more simultaneous failures of the AP1000 passive safety systems during large- and small-break loss-of-coolant accidents, including station blackout and cold shutdown conditions. The experiments run in APEX-AP1000 confirm significant liquid entrainment and carryover of water to the automatic depressurization system (ADS) during and after actuation of the fourth-stage (ADS4) valves. These processes are important as thermal-hydraulic codes used to analyze the AP1000 design must adequately predict or bound upper plenum and hot leg entrainment. In addition, the tests show that failure of ADS4 valves on the non-pressurizer side of the plant results in a greater delay in in-containment refueling water storage tank (IRWST) injection than failure of ADS4 valves on the pressurizer side of the plant; while failure of two out of four ADS4 valves on the non-pressurizer side of the plant produces low two-phase mixture levels in the core during certain simulated vessel injection line and cold leg breaks.