Expert system for the CPCS-initiated trip analysis

Conference · Tue Jan 01 00:00:00 EST 1991 · Transactions of the American Nuclear Society; (United States)

OSTI ID:5978008

Sohn, Sedo; Im, Inyoung; Kuh, Jungeui ^[1]

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1. Korea Atomic Energy Research Inst., Taejon (Korea)

In Yonggwang nuclear units 3 and 4, the core protection calculator system (CPCS) performs various protection logics against many transients and certain accidents. The CPCS is a real-time computer system calculating the departure from nucleate boiling ratio (DNBR), and local power density, and other protection logics. It takes process variables such as neutron flux, hot-leg temperature, cold-leg temperature, control element assembly positions, and reactor coolant pump shaft speed. Since the CPCS protection logics are quite complex, it is difficult for an operator to tell immediately which parameter is the major cause of the reactor trip. Thus, whenever the reactor trip signal is generated, the process input variables and calculated results, including selected intermediate variables, are frozen in the specified computer memory for later analysis. These frozen variables are called the trip buffer. Analysis of the trip buffer requires an expert in the CPCS and related documents containing algorithms and a data base for algorithms. The Trip Buffer Analysis Program (TBAP) is an expert system that pinpoints the causes of the CPCS initiated reactor trip, thus relieving the operator from the burden of analyzing the trip buffer.

• OSTI ID: 5978008
• Report Number(s): CONF-910603-; CODEN: TANSA
• Journal Information: Transactions of the American Nuclear Society; (United States), Vol. 63; Conference: Annual meeting of the American Nuclear Society (ANS), Orlando, FL (United States), 2-6 Jun 1991; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
PWR TYPE REACTORS
REACTOR PROTECTION SYSTEMS
DIAGNOSTIC TECHNIQUES
EXPERT SYSTEMS
ALGORITHMS
DATA BASE MANAGEMENT
DEPARTURE NUCLEATE BOILING
DESIGN
DOCUMENTATION
NEUTRON FLUX
PERFORMANCE
POWER DENSITY
PRIMARY COOLANT CIRCUITS
PROGRAMMING
REACTOR ACCIDENTS
REACTOR OPERATORS
REAL TIME SYSTEMS
SCRAM
TEMPERATURE CONTROL
TRANSIENTS
VALIDATION
ACCIDENTS
BOILING
CONTROL
COOLING SYSTEMS
ENRICHED URANIUM REACTORS
MANAGEMENT
MATHEMATICAL LOGIC
NUCLEATE BOILING
PERSONNEL
PHASE TRANSFORMATIONS
POWER REACTORS
RADIATION FLUX
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR SHUTDOWN
REACTORS
SHUTDOWN
TESTING
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled