Comparison of realistic large break LOCA analyses of a 3-loop westinghouse plant using response surface and statistical sampling techniques

Proceedings of the International Conference on Nuclear Engineering (ICONE12)

Volumn 3, Issue , 2004, Pages 811-818

  Muftuoglu, K ^a   Ohkawa, K ^a   Frepoli, C ^a   Nissley, M ^a  

^a WESTINGHOUSE ELECTRIC COMPANY   (United States)

Author keywords

[No Author keywords available]

Indexed keywords

ACCIDENT PREVENTION; COOLING SYSTEMS; REACTOR CORES; REGULATORY COMPLIANCE; SAMPLING; STATISTICAL METHODS;

APPLICABILITY; EMERGENCY CORE COOLING SYSTEM (ECCS); STATISTICAL SAMPLING TECHNIQUES;

NUCLEAR POWER PLANTS;

EID: 10644233900     PISSN: None     EISSN: None     Source Type: Conference Proceeding    
DOI: 10.1115/icone12-49499     Document Type: Conference Paper

References (10)

1. Bajorek, S. M., et al., 1998, "Code Qualification Document for Best Estimate LOCA Analysis," Technical Report WCAP-12945-NP-A, Volume 1, Revision 2, and Volumes 2 through 5, Revision 1, Westinghouse Electric Company, Pittsburgh, PA.
2. Boyack, B. E., et al., 1989, "Quantifying Reactor Safety Margins," NUREG/CR-5249.
3. Glaeser, H., Hofer, E., Kloos, M., and Skorek, T., 1998, "GRS Analyses for CSNI Uncertainty Methods Study (UMS)," in CSNI Report NEA/CSNI/R(97)35, Volume 1, June 1998.
4. Glaeser, H., 2000, "Uncertainty Evaluation of Thermal-Hydraulic Code Results," Proc. Int. Mtg. on "Best-Estimate" Methods in Nucl. Installation Safety Analysis (BE-2000) Washington D.C., November 2000.
5. Liles, D. R., et al., 1981, "TRAC-PD2, An Advanced Best Estimate Computer Program for Pressurized Water Reactor Loss-of-Coolant Accident Analysis," NUREG/CR-2054, Los Alamos National Laboratories, NM.
6. Nissley, M. E., Frepoli, C., Ohkawa, K., and Muftuoglu, K., 2003, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," Technical Report WCAP-16009-NP, Westinghouse Electric Company, Pittsburgh, PA.
7. Thurgood, M. J., et al., 1980, "COBRA-TF Development," 8th Water Reactor Safety Information Meeting.
8. Thurgood, M. J., et al., 1983, "COBRA/TRAC - A Thermal-Hydraulics Code for Transient Analysis of Nuclear Reactor Vessels and Primary Coolant Systems," NUREG/CR-3046.
9. Wickelt, T. (editor), et al., 1998, "Report of the Uncertainty Methods Study for Advanced Best Estimate Thermal Hydraulic Code Applications," NEA/CSNI/R(97)35.
10. Wilks, S. S., 1941, "Determination of Sample Sizes for Setting Tolerance Limits," The Annals of Mathematical Statistics, Vol. 12, pp. 91-96.