Kuosheng BWR/6 recirculation pump trip transient analysis with the RETRAN02/MOD5 code

Conference · Wed Jan 01 04:00:00 EST 1992 · Transactions of the American Nuclear Society; (United States)

OSTI ID:6983593

Wang, J R; Shih, C ^[1]

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1. National Tsing-Hua Univ., Hsinchu (Taiwan, Province of China)

A recirculation pump trip (RPT) event results in a reduction in recirculation flow, which reduces the core coolant flow rate. A reduction in core flow results in an increase in core void fraction and hence a decrease in core power due to negative void reactivity feedback. Although this category of events is less severe than others and generally considered as nonlimiting, core instability still may occur such as that at LaSalle on March 9, 1988. This paper focuses on the RPT transient analysis of Kuosheng Nuclear Power Plant (KNPP), which has two units of General Electric-designed boiling water reactor (BWR)/6 with rated core thermal power of 2894 MW and rated core flow of 10645 kg/s (23472 lb[sub m]/s). The approach to investigating the RPT transient of KNPP consists of two steps. The first step is to develop a plant-specific model using the RETRAN02/MOD5 code. In this step, various plant-specific information, including design documentation, drawings, safety analysis reports, and other information supplied by vendors were collected for model development. The RPT startup test at 68% power was used for system model benchmarking to ensure the adequacy of this model and identify several sensitive parameters. The second step is to assess whether similar power oscillation phenomena may occur at KNPP because of an RPT with isolated feedwater heater event. Two transient analyses (with or without reactor scram) of the KNPP RPT with isolated feedwater heater were investigated.

• OSTI ID: 6983593
• Report Number(s): CONF-921102--
• Conference Information: Journal Name: Transactions of the American Nuclear Society; (United States) Journal Volume: 66
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100* -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
ACCIDENTS
BWR TYPE REACTORS
COMPUTER CODES
COOLING SYSTEMS
ENRICHED URANIUM REACTORS
KUOSHENG-1 REACTOR
KUOSHENG-2 REACTOR
LOSS OF FLOW
OSCILLATIONS
POWER DISTRIBUTION
POWER REACTORS
PRIMARY COOLANT CIRCUITS
PUMPS
R CODES
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR STABILITY
REACTORS
STABILITY
THERMAL REACTORS
TRANSIENTS
VOID FRACTION
WATER COOLED REACTORS
WATER MODERATED REACTORS