Operational performance risk assessment in support of a supervisory control system - 251

Conference · Thu Jun 15 04:00:00 EDT 2017

OSTI ID:23035366

Guler, A.; Muhlheim, M.; Cetiner, S. ^[1]; Denning, R. ^[2]

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1. Oak Ridge National Laboratory (United States)
2. Research Consultant (United States)

A supervisory control system is being developed for multiunit advanced small modular reactors to minimize human interventions during normal and abnormal operations. In the supervisory control system, control action decisions are made based on a probabilistic risk assessment that employs event trees and fault trees. Although traditional probabilistic risk assessment tools are implemented, their scope is extended to normal operations, and the application is reversed to assess the success of non-safety related systems and to enable continued operation of the plant. This extended probabilistic risk assessment approach is called operational performance risk assessment (OPRA). OPRA helps to identify available paths, combine control actions for maintaining plant conditions within operational limits, and to quantify the likelihood of success of these operational trajectories to optimize the selection of alternative actions without activating reactor protection system. In this paper, a case study of OPRA in a supervisory control system is demonstrated for the Advanced Liquid Metal Reactor (ALMR) Power Reactor Inherently Safe Module (PRISM) design, specifically the power conversion system. The scenario investigated involved a condition in which the feedwater control valve that was observed to be drifting to the closed position. Alternative plant configurations that would allow the plant to continue to operate at full or reduced power were identified using OPRA. Dynamic analyses were performed with a thermal-hydraulic model of the ALMR PRISM system using Modelica to evaluate the magnitude of safety margins. Successful recovery paths for the selected scenario were identified and quantified using the supervisory control system. (authors)

• Research Organization: American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 23035366
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND TECHNOLOGY
CONFIGURATION
DESIGN
FAULT TREE ANALYSIS
FEEDWATER
LIQUID METAL COOLED REACTORS
PERFORMANCE
POWER REACTORS
PROBABILISTIC ESTIMATION
REACTOR CONTROL SYSTEMS
REACTOR OPERATION
REACTOR PROTECTION SYSTEMS
RISK ASSESSMENT
SAFETY MARGINS
SMALL MODULAR REACTORS
THERMAL HYDRAULICS
VALVES