Analysis of core damage frequency from internal events: Surry, Unit 1
Technical Report
·
OSTI ID:6660464
This document contains the accident sequence analyses for Surry, Unit 1; one of the reference plants being examined as part of the NUREG-1150 effort by the Nuclear Regulatory Commission (NRC). NUREG-1150 will document the risk of a selected group of nuclear power plants. As part of that work, this report contains the overall core damage frequency estimate for Surry, Unit 1, and the accompanying plant damage state frequencies. Sensitivity and uncertainty analyses provide additional insights regarding the dominant contributors to the Surry core damage frequency estimate. The numerical results are driven to some degree by modeling assumptions and data selection for issues such as reactor coolant pump seal LOCAs, common cause failure probabilities, and plant response to station blackout and loss of electrical bust initiators. The sensitivity studies explore the impact of alternate theories and data on these issues.
- Research Organization:
- Sandia National Labs., Albuquerque, NM (USA); Nuclear Regulatory Commission, Washington, DC (USA). Div. of Reactor Systems Safety
- DOE Contract Number:
- AC04-76DP00789
- OSTI ID:
- 6660464
- Report Number(s):
- NUREG/CR-4550-Vol.3; SAND-86-2084-Vol.3; ON: TI87007079
- Country of Publication:
- United States
- Language:
- English
Similar Records
Analysis of core damage frequency from internal events: Sequoyah, Unit 1
Aperture cards: Sequoyah Nuclear Plant (Engineering Materials)
Analysis of core damage frequency, Surry, Unit 1 internal events appendices
Technical Report
·
Sat Jan 31 23:00:00 EST 1987
·
OSTI ID:6570407
Aperture cards: Sequoyah Nuclear Plant (Engineering Materials)
Miscellaneous
·
Sat Jan 31 23:00:00 EST 1987
·
OSTI ID:5717407
Analysis of core damage frequency, Surry, Unit 1 internal events appendices
Technical Report
·
Sat Mar 31 23:00:00 EST 1990
·
OSTI ID:7146808
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
DAMAGE
DATA ACQUISITION
DATA BASE MANAGEMENT
DESIGN
ENERGY TRANSFER
ENGINEERING
ENRICHED URANIUM REACTORS
FAILURE MODE ANALYSIS
FAULT TREE ANALYSIS
FLUID MECHANICS
HEAT TRANSFER
HUMAN FACTORS ENGINEERING
HYDRAULICS
LOSS OF COOLANT
MANAGEMENT
MATHEMATICAL MODELS
MATHEMATICS
MECHANICS
MELTDOWN
OPERATION
POWER REACTORS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR CORES
REACTOR OPERATION
REACTOR SHUTDOWN
REACTORS
RISK ASSESSMENT
SCRAM
SENSITIVITY ANALYSIS
SHUTDOWNS
STATISTICAL MODELS
STATISTICS
SURRY-1 REACTOR
SYSTEM FAILURE ANALYSIS
SYSTEMS ANALYSIS
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
DAMAGE
DATA ACQUISITION
DATA BASE MANAGEMENT
DESIGN
ENERGY TRANSFER
ENGINEERING
ENRICHED URANIUM REACTORS
FAILURE MODE ANALYSIS
FAULT TREE ANALYSIS
FLUID MECHANICS
HEAT TRANSFER
HUMAN FACTORS ENGINEERING
HYDRAULICS
LOSS OF COOLANT
MANAGEMENT
MATHEMATICAL MODELS
MATHEMATICS
MECHANICS
MELTDOWN
OPERATION
POWER REACTORS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR CORES
REACTOR OPERATION
REACTOR SHUTDOWN
REACTORS
RISK ASSESSMENT
SCRAM
SENSITIVITY ANALYSIS
SHUTDOWNS
STATISTICAL MODELS
STATISTICS
SURRY-1 REACTOR
SYSTEM FAILURE ANALYSIS
SYSTEMS ANALYSIS
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS