Statistically based uncertainty analysis for ranking of component importance in the thermal-hydraulic safety analysis of the Advanced Neutron Source Reactor
The Analytic Hierarchy Process (AHP) has been used to help determine the importance of components and phenomena in thermal-hydraulic safety analyses of nuclear reactors. The AHP results are based, in part on expert opinion. Therefore, it is prudent to evaluate the uncertainty of the AHP ranks of importance. Prior applications have addressed uncertainty with experimental data comparisons and bounding sensitivity calculations. These methods work well when a sufficient experimental data base exists to justify the comparisons. However, in the case of limited or no experimental data the size of the uncertainty is normally made conservatively large. Accordingly, the author has taken another approach, that of performing a statistically based uncertainty analysis. The new work is based on prior evaluations of the importance of components and phenomena in the thermal-hydraulic safety analysis of the Advanced Neutron Source Reactor (ANSR), a new facility now in the design phase. The uncertainty during large break loss of coolant, and decay heat removal scenarios is estimated by assigning a probability distribution function (pdf) to the potential error in the initial expert estimates of pair-wise importance between the components. Using a Monte Carlo sampling technique, the error pdfs are propagated through the AHP software solutions to determine a pdf of uncertainty in the system wide importance of each component. To enhance the generality of the results, study of one other problem having different number of elements is reported, as are the effects of a larger assumed pdf error in the expert ranks. Validation of the Monte Carlo sample size and repeatability are also documented.
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls, ID (United States)
- Sponsoring Organization:
- USDOE; USDOE, Washington, DC (United States)
- DOE Contract Number:
- AC07-76ID01570
- OSTI ID:
- 5485664
- Report Number(s):
- EGG-NE-10078; ON: DE92012311
- Country of Publication:
- United States
- Language:
- English
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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
99 GENERAL AND MISCELLANEOUS//MATHEMATICS, COMPUTING, AND INFORMATION SCIENCE
RESEARCH REACTORS
SAFETY ANALYSIS
AFTER-HEAT REMOVAL
DATA COVARIANCES
DISTRIBUTION FUNCTIONS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
MONTE CARLO METHOD
PROBABILITY
REACTOR COMPONENTS
REACTOR SAFETY
STATISTICAL DATA
ACCIDENTS
DATA
ENERGY TRANSFER
FLUID MECHANICS
FUNCTIONS
INFORMATION
MECHANICS
NUMERICAL DATA
REACTOR ACCIDENTS
REACTORS
REMOVAL
RESEARCH AND TEST REACTORS
SAFETY
220900* - Nuclear Reactor Technology- Reactor Safety
220600 - Nuclear Reactor Technology- Research
Test & Experimental Reactors
990200 - Mathematics & Computers