Steam generator model validation and advanced feedwater control system design for the Maanshan PWR (pressurized water reactor)

Technical Report · Fri Sep 01 00:00:00 EDT 1989

OSTI ID:5679785

Gaydos, K A ^[1]

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1. Westinghouse Electric Corp., Pittsburgh, PA (USA). Nuclear and Advanced Technology Div.

This is the Final Report of Project RP2126-7, a cooperative effort between Taiwan Power Company, EPRI, and Westinghouse which resulted in two major outcomes: the actual validation of the key low power transient response characteristics of a practical dynamic computer model of the vertical natural circulation U-tube steam generators at the Maanshan pressurized water reactor in Taiwan, and the adaptation of the Westinghouse Advanced Digital Feedwater Control System (ADFCS) control algorithms functional design for application at Maanshan. The nonlinear multinode steam generator model is first outlined. A specially designed test program conducted at Maanshan is described. These tests isolate the key dynamic responses of the steam generator to open loop perturbations in boundary conditions at low power. The validation of the model with test data is then described. Results of comparisons between the measured and the measured versus predicted comparisons of the well-known steam generator level shrink/swell phenomenon show good agreement. The technical features and benefits of the Maanshan ADFCS control algorithms functional design are summarized. This improved functional strategy has proven to provide effective feedwater control over the full range of plant operation at operating nuclear station in Europe. 10 refs., 35 figs.

• Research Organization: Electric Power Research Inst., Palo Alto, CA (USA); Westinghouse Electric Corp., Pittsburgh, PA (USA). Nuclear and Advanced Technology Div.
• Sponsoring Organization: EPRI
• OSTI ID: 5679785
• Report Number(s): EPRI-NP-6506; TRN: 89-027757
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
99 GENERAL AND MISCELLANEOUS//MATHEMATICS, COMPUTING, AND INFORMATION SCIENCE
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
FEEDWATER
COMPUTERIZED CONTROL SYSTEMS
REACTOR CONTROL SYSTEMS
PWR TYPE REACTORS
ALGORITHMS
COMPUTERIZED SIMULATION
HEAT TRANSFER
HYDRAULICS
PROGRESS REPORT
REACTOR COOLING SYSTEMS
STEAM GENERATORS
TAIWAN
TRANSIENTS
ASIA
BOILERS
CONTROL SYSTEMS
COOLING SYSTEMS
DOCUMENT TYPES
ENERGY SYSTEMS
ENERGY TRANSFER
FLUID MECHANICS
HYDROGEN COMPOUNDS
ISLANDS
MATHEMATICAL LOGIC
MECHANICS
OXYGEN COMPOUNDS
REACTOR COMPONENTS
REACTORS
SIMULATION
VAPOR GENERATORS
WATER
WATER COOLED REACTORS
WATER MODERATED REACTORS
220400* - Nuclear Reactor Technology- Control Systems
990220 - Computers
Computerized Models
& Computer Programs- (1987-1989)
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled