Evolution of PRA methodology and insights since WASH-1400
Conference
·
· Trans. Am. Nucl. Soc.; (United States)
OSTI ID:7124396
The US Nuclear Regulatory Commission (NRC) is preparing NUREG-1150 to examine the current perception of risk from a selected group of nuclear power plants. In support of NUREG-1150, Sandia National Laboratories (SNL) has directed the production of Level 1 Probabilistic Risk Assessments (PRAs) for the Surry, Sequoyah, Peach Bottom, and Grand Gulf nuclear power plants; additional studies are planned. The first four plants have been studied previously in either WASH-1400 or RSSMAP. The more recent studies suggest significant changes in perception of dominant accident sequences. In this paper the authors examine the changes in their perception of the likelihood of severe core damage accidents, in terms of both changes in PRA methodology and changes to the plants as a result of evolving regulations.
- Research Organization:
- Sandia National Labs., Albuquerque, NM
- OSTI ID:
- 7124396
- Report Number(s):
- CONF-861102-
- Conference Information:
- Journal Name: Trans. Am. Nucl. Soc.; (United States) Journal Volume: 53
- Country of Publication:
- United States
- Language:
- English
Similar Records
Evolution of PRA methodology and insights since WASH-1400
Smart approach to level-1 probabilistic risk assessment
Smart approach to Level 1 probabilistic risk assessment
Conference
·
Tue Dec 31 23:00:00 EST 1985
·
OSTI ID:5291083
Smart approach to level-1 probabilistic risk assessment
Conference
·
Tue Dec 31 23:00:00 EST 1985
· Trans. Am. Nucl. Soc.; (United States)
·
OSTI ID:7124426
Smart approach to Level 1 probabilistic risk assessment
Conference
·
Tue Dec 31 23:00:00 EST 1985
·
OSTI ID:5358038
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100 -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
AFTER-HEAT REMOVAL
AVAILABILITY
BWR TYPE REACTORS
CONTAINMENT
CONTAINMENT SYSTEMS
CONTROL EQUIPMENT
CONTROL ROD DRIVES
ECCS
ENGINEERED SAFETY SYSTEMS
ENRICHED URANIUM REACTORS
EQUIPMENT
FAILURES
FLOW REGULATORS
GRAND GULF-1 REACTOR
LOSS OF COOLANT
NATIONAL ORGANIZATIONS
OUTAGES
PEACH BOTTOM-2 REACTOR
POWER REACTORS
PROBABILITY
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR CORE DISRUPTION
REACTOR PROTECTION SYSTEMS
REACTOR SHUTDOWN
REACTORS
RELIABILITY
REMOVAL
RISK ASSESSMENT
SANDIA LABORATORIES
SEQUOYAH-1 REACTOR
SHUTDOWNS
SURRY-1 REACTOR
THERMAL REACTORS
US AEC
US DOE
US ERDA
US NRC
US ORGANIZATIONS
VALVES
WATER COOLED REACTORS
WATER MODERATED REACTORS
210100 -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
AFTER-HEAT REMOVAL
AVAILABILITY
BWR TYPE REACTORS
CONTAINMENT
CONTAINMENT SYSTEMS
CONTROL EQUIPMENT
CONTROL ROD DRIVES
ECCS
ENGINEERED SAFETY SYSTEMS
ENRICHED URANIUM REACTORS
EQUIPMENT
FAILURES
FLOW REGULATORS
GRAND GULF-1 REACTOR
LOSS OF COOLANT
NATIONAL ORGANIZATIONS
OUTAGES
PEACH BOTTOM-2 REACTOR
POWER REACTORS
PROBABILITY
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR CORE DISRUPTION
REACTOR PROTECTION SYSTEMS
REACTOR SHUTDOWN
REACTORS
RELIABILITY
REMOVAL
RISK ASSESSMENT
SANDIA LABORATORIES
SEQUOYAH-1 REACTOR
SHUTDOWNS
SURRY-1 REACTOR
THERMAL REACTORS
US AEC
US DOE
US ERDA
US NRC
US ORGANIZATIONS
VALVES
WATER COOLED REACTORS
WATER MODERATED REACTORS