Nuclear power plant digital system PRA pilot study with the dynamic flow-graph methodology

Conference · Sat Jul 01 04:00:00 EDT 2006

OSTI ID:22030179

Yau, M; Motamed, M; Guarro, S ^[1]

---

1. ASCA, Inc., 1720 South Catalina Avenue, Redondo Beach, CA 90277 (United States)

Current Probabilistic Risk Assessment (PRA) methodology is well established in analyzing hardware and some of the key human interactions. However processes for analyzing the software functions of digital systems within a plant PRA framework, and accounting for the digital system contribution to the overall risk are not generally available nor are they well understood and established. A recent study reviewed a number of methodologies that have potential applicability to modeling and analyzing digital systems within a PRA framework. This study identified the Dynamic Flow-graph Methodology (DFM) and the Markov Methodology as the most promising tools. As a result of this study, a task was defined under the framework of a collaborative agreement between the U.S. Nuclear Regulatory Commission (NRC) and the Ohio State Univ. (OSU). The objective of this task is to set up benchmark systems representative of digital systems used in nuclear power plants and to evaluate DFM and the Markov methodology with these benchmark systems. The first benchmark system is a typical Pressurized Water Reactor (PWR) Steam Generator (SG) Feedwater System (FWS) level control system based on an earlier ASCA work with the U.S. NRC 2, upgraded with modern control laws. ASCA, Inc. is currently under contract to OSU to apply DFM to this benchmark system. The goal is to investigate the feasibility of using DFM to analyze and quantify digital system risk, and to integrate the DFM analytical results back into the plant event tree/fault tree PRA model. (authors)

• Research Organization: American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 22030179
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
97 MATHEMATICS AND COMPUTING
BENCHMARKS
COMPUTER CODES
COMPUTERIZED SIMULATION
DIGITAL SYSTEMS
FAULT TREE ANALYSIS
FEEDWATER
GRAPH THEORY
MARKOV PROCESS
NUCLEAR POWER PLANTS
PROBABILISTIC ESTIMATION
PWR TYPE REACTORS
REACTOR CONTROL SYSTEMS
RISK ASSESSMENT
STEAM GENERATORS