Using the PC software PCTRAN for IPE core-melt sequence analysis at TMI-1

Authors: Cliff Po, Li-Chi (GPU Nuclear Corp., Parsippany, NJ (United States))
 
Abstract: In preparing for the individual plant examination (IPE) of Three Mile Island Unit I (TMI-1), two events that contribute significantly to the level-I core damage risk are station black- out (SBO) and steam generator tube rupture (SGTR).^For an SBO, it is assumed that both the off-site power and on-site diesel generators fail to supply alternating-current power for the plant systems.^Timing to core uncovery is important because either the off-site or on-site power may be recovered during the course of the event.^For an SGTR, proper operator action may mitigate the consequences and prevent core damage.^Traditional system transient analysis codes are not practical to conduct transient prediction of these events because of their long running time and the large number of cases to be analyzed.^The personal computer-based plant analyzer code PCTRAN was used because of its simplicity and fast turn-around time.^The code and the Babcock and Wilcox-designed pressurized water reactor plant, model have been previously verified against real plant data.^With the exception of the large-break loss-of-coolant accident, for which the sudden momentum effect is not accounted for by the simple mass and energy balance equations, PCTRAN can generally reproduce operational transients and slowly varying accidents with reasonable accuracy in a faster-than-real-time mode.
Publication Date: 01 Jan 1993
Report numbers: CONF-930601--
Resource Type: Conference
Resource Specific Type: Journal Article
Resource Relation: Transactions of the American Nuclear Society ; Vol/Issue: 68; American Nuclear Society (ANS) annual meeting; 20-24 Jun 1993; San Diego, CA (United States)
Country of Publication: United States
Language: English
Keywords relating to this report:
-- NUCLEAR REACTOR TECHNOLOGY-- REACTOR SAFETY
-- POWER REACTORS, NONBREEDING, LIGHT-WATER MODERATED, NONBOILING WATER COOLED
ECCS-- COMPUTERIZED SIMULATION
P CODES
PERSONAL COMPUTERS
REACTOR PROTECTION SYSTEMS
REACTOR SAFETY
THREE MILE ISLAND-1 REACTOR-- SAFETY ANALYSIS
Related subjects:
COMPUTER CODES
COMPUTERS
DIGITAL COMPUTERS
ENGINEERED SAFETY SYSTEMS
ENRICHED URANIUM REACTORS
MICROCOMPUTERS
POWER REACTORS
PWR TYPE REACTORS
REACTOR PROTECTION SYSTEMS
REACTORS
SAFETY
SIMULATION
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS