Analysis of loss of off-site power with a PWR at shutdown
In many probabilistic risk assessments (PRAs), loss of offsite power (LOOP) when a nuclear power plant is operating was found to be a significant contributor to core damage. The purpose of this study is to provide an analysis of a LOOP event that occurs while a pressurized water reactor (PWR) is shut down. The importance of such an analysis was recognized as part of a study to evaluate the core damage frequency due to a loss of decay heat removal (DHR) capability during an outage. When a PWR is in a shutdown condition, there are relatively few technical specification requirements on the operability of safety systems. In fact, some safety systems are intentionally disabled, i.e., the safety injection system and nonoperating charging pumps. Another problem when the reactor is shut down is that the reactor coolant system (RCS) may be partially drained and the steam generators may be unavailable. To determine the time available for operator actions, given that a LOOP occurs during shutdown and the DHR capability is lost, a simple thermal model has been developed. Similar calculations have been performed for other phases of refueling and maintenance outages. A total core damage frequency due to LOOP while the plant is in shutdown has been calculated to be 5.9 x 10/sup -6//yr. This is approximately twice the core damage frequency calculated for LOOP when the plant is at power.
- Research Organization:
- Brookhaven National Lab., Upton, NY (USA)
- OSTI ID:
- 6897942
- Report Number(s):
- CONF-8711195-
- Journal Information:
- Trans. Am. Nucl. Soc.; (United States), Journal Name: Trans. Am. Nucl. Soc.; (United States) Vol. 55; ISSN TANSA
- Country of Publication:
- United States
- Language:
- English
Similar Records
Simulation of loss-of-decay heat removal with MAAP 3. 0B
Analysis of air ingression into the core region during hypothetical PWR shutdown accidents
Related Subjects
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
AC SYSTEMS
ACCIDENTS
AFTER-HEAT REMOVAL
COOLING SYSTEMS
ENERGY SYSTEMS
MAINTENANCE
OUTAGES
POWER SYSTEMS
PROBABILITY
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR CORE DISRUPTION
REACTOR FUELING
REACTOR MAINTENANCE
REACTOR SAFETY
REACTOR SHUTDOWN
REACTORS
REMOVAL
RHR SYSTEMS
RISK ASSESSMENT
SAFETY
SHUTDOWNS
WATER COOLED REACTORS
WATER MODERATED REACTORS