A probabilistic method for evaluating reactivity feedbacks and its application to EBR-II
The probability that reactivity feedbacks fail to prevent damage is computed by propagating data and modeling uncertainties through transient calculations, with these uncertainties being constrained by experimental evidence. Screening processes are used to identify the most important parameters and accident initiators. The response surface method is used to facilitate the error propagation and a Monte Carlo rejection technique is used to force the parameter variations to be consistent with the observed distribution of experimental quantities. The reliability of the failure probability estimates is evaluated. This process is applied to ATWS events in the PRA for the EBR-II reactor. The loss-of-normal-power (LONP), loss-of-flow and transient overpower accidents without scram were found to warrant detailed analysis and a complete analysis has been made for the first of these. Six parameters are primarily responsible for the LONP outcome variations. The conditional probability of minor core damage from LONP without scram is 1.2 {times} 10{sup {minus}2}. The uncertainty in this estimate is a factor of 2. This damage estimate would be an order of magnitude higher if experimental information about feedbacks in EBR-II was not used. the conditional probability of major core damage from LONP without scram is <10{sup {minus}6}. 20 refs., 1 fig., 3 tabs.
- Research Organization:
- Argonne National Lab., Idaho Falls, ID (USA)
- Sponsoring Organization:
- DOE/NE
- DOE Contract Number:
- W-31109-ENG-38
- OSTI ID:
- 6233661
- Report Number(s):
- CONF-910414-6; ON: DE91007211; TRN: 91-008331
- Resource Relation:
- Conference: International topical meeting on advances in mathematics, computations and reactor physics, Pittsburgh, PA (USA), 28 Apr - 2 May 1991
- Country of Publication:
- United States
- Language:
- English
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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
EBR-2 REACTOR
REACTOR CORE DISRUPTION
REACTOR KINETICS
RISK ASSESSMENT
REACTIVITY
AFTER-HEAT REMOVAL
ANL
ATWS
BLACKOUTS
CONTROL ELEMENTS
DEFORMATION
EFFICIENCY
ERRORS
EXPERIMENTAL DATA
FAILURES
FISSION PRODUCT RELEASE
FUEL-CLADDING INTERACTIONS
LOSS OF FLOW
MATHEMATICAL MODELS
MITIGATION
MONTE CARLO METHOD
NATURAL CONVECTION
P CODES
PRIMARY COOLANT CIRCUITS
PROBABILITY
PUMPS
REACTOR ACCIDENTS
REACTOR OPERATION
REACTOR SAFETY
RELIABILITY
RESPONSE FUNCTIONS
S CODES
SODIUM
TRANSIENT OVERPOWER ACCIDENTS
TRANSIENTS
ACCIDENTS
ALKALI METALS
BREEDER REACTORS
COMPUTER CODES
CONVECTION
COOLING SYSTEMS
DATA
ELEMENTS
ENERGY SYSTEMS
ENERGY TRANSFER
EPITHERMAL REACTORS
EXPERIMENTAL REACTORS
FAST REACTORS
FBR TYPE REACTORS
FUNCTIONS
HEAT TRANSFER
INFORMATION
KINETICS
LIQUID METAL COOLED REACTORS
LMFBR TYPE REACTORS
MASS TRANSFER
METALS
NATIONAL ORGANIZATIONS
NUMERICAL DATA
OPERATION
POWER REACTORS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTORS
REMOVAL
RESEARCH AND TEST REACTORS
SAFETY
SODIUM COOLED REACTORS
US AEC
US DOE
US ERDA
US ORGANIZATIONS
220900* - Nuclear Reactor Technology- Reactor Safety
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