Evaluating code uncertainty-I: Using the csau method for uncertainty analysis of a two-loop pwr sbloca

Nuclear Technology

Volumn 126, Issue 2, 1999, Pages 170-185

  Prosek, Andrej ^a   Mavko, Borut ^a  

^a JO EF STEFAN INSTITUTE   (Slovenia)

Author keywords

Thermal hydraulics; Transient and accident analysis; Uncertainty analysis

Indexed keywords

COMPUTER AIDED ANALYSIS; COMPUTER SOFTWARE; EXTRAPOLATION; HYDRAULICS; PRESSURIZED WATER REACTORS;

SOFTWARE PACKAGE RELAP/MOD; THERMAL HYDRAULICS; UNCERTAINTY ANALYSIS;

LOSS OF COOLANT ACCIDENTS;

EID: 0344541698     PISSN: 00295450     EISSN: None     Source Type: Journal    
DOI: 10.13182/NT99-A2965     Document Type: Article

References (45)

1. "Emergency Core Cooling Systems, Revisions to Acceptance Criteria," 53FR180, Federal Register (Sep. 16, 1988).
2. "Domestic Licensing of Production and Utilization Facilities," Title 10, Part 50, Appendix K, Code of Federal Regulations (Jan. 1992)
3. B. BOYACK, R. DUFFEY, P. GRIFFITH, G. LELLOUCHE, S. LEVY, U. ROHATGI, G. WILSON, W. WULFF, and N. ZUBER, "Quantifying Reactor Safety Margins, Application of Code Scaling, Applicability, and Uncertainty Evaluation Methodology to a Large-Break, Loss-of-Coolant Accident," NUREG/CR-5249, U.S. Nuclear Regulatory Commission (1989).
4. H. GLAESER, E. HOFER, E. CHONJACKI, A. QUNSY, and C. RENAULT, "Mathematical Techniques for Uncertainty and Sensitivity Analysis," P roc. Validation of Systems Transients Analysis Codes, Hilton Head, South Carolina, August 13-18, 1995, p. 75, FED-Vol. 223, American Society of Mechanical Engineers (1995).
5. H. GLAESER and R. POCHARD, "Review on Uncertainty Methods for Thermal Hydraulic Computer Codes," Proc. Int. Conf. New Trends in Nuclear System Thermalhydraulics, Pisa, Italy, May 30-June 2, 1994, Vol. 1, p. 447, Edizioni ETS (1994).
6. F. D'AURIA, N. DEBRECIN, and G. M. GALASSI, "Outline of the Uncertainty Methodology Based on Accuracy Extrapolation," M/cJ. Technol, 109, 21 (1995).
7. M. REOCREUX, "Safety Analysis and Best Estimate Codes," Proc. 4th Mtg. Nuclear Energy in Central Europe, Bled, Slovenia, September 7-10, 1997, p. 9, Nuclear Society of Slovenia (1997).
8. A. QUINSY, B. BRUN, and F. DE CRÉCY, "The Adjoint Sensitivity Method, A Contribution to the Code Uncertainty Evaluation," J. Nucl. Eng. Des., 149, 357 (1994).
9. TECHNICAL PROGRAM GROUP: B. E. BOYACK, I. CATTON, R. B. DUFFEY, P. GRIFFITH, K. R. KATSMA, G. S. LELLOUCHE, S. LEVY, U. S. ROHATGI, G. E. WILSON, W. WULFF, and N. ZUBER, "Quantifying Reactor Safety Margin, Parts 1 to 6," J. Nucl. Eng. Des., 119, l (1990).
10. M. G. ORTIZ and L. S. GHAN, "Uncertainty Analysis of Minimum Vessel Liquid Inventory During a Small-Break LOCA in a B&W Plant-An Application of the CSAU Methodology Using the RELAP5/MOD3 Computer Code," NUREG/CR5818, EGG-2665, Idaho National Engineering Laboratory (1992). H.H. GLAESER, "Validation and Uncertainty Analysis of the ATHLET Thermal-Hydraulic Computer Code," Proc. 2nd Regional Mtg. Nuclear Energy in Central Europe, Portoroz, Slovenia, September 11-14, 1995, p. 591, Nuclear Society of Slovenia (1995).
11. S. R. PARK, W. P. BAEK, S. H. CHANG, and B. H. LEE, "Development of an Uncertainty Quantification Method of the Best Estimate Large LOCA Analysis," J. Nucl Eng. Des., 135, 367 (1992).
12. B. MAVKO, A. STRITAR, and A. PROSEK, "Application of Code Scaling, Applicability and Uncertainty Methodology to Large Break LOCA Analysis of Two-Loop PWR," J. Nucl. Eng. Des., 143, 95 (1993).
13. E. F. HASKIN, B. D. BEVAN, and C. DING, "Efficient Uncertainty Analyses Using Fast Probability Integration," J. Nucl. Eng. Des., 166, 225 (1996).
14. W. S. YEUNG, R. C. HARVEY, J. SHIRKOV, R. K. SUNDARAM, and S. SARDY, "Application of the CSAU Methodology to BETHSY SBLOCA Test 9.1b Using RELAP5/ MOD3.1," presented at RELAP5 International Users Seminar, Dallas, Texas, March 17-21, 1996.
15. G. E. WILSON, B. E. BOYACK, I. CATTON, R. B. DUFFEY, P. GRIFFITH, K. R. KATSMA, G. S. LELLOUCHE, S. LEVY, U. S. ROHATGI, W. WULFF, and N. ZUBER, "TPG Response to the Foregoing Letters-to-the-Editor," J. Nucl. Eng. Des., 132,431 (1992).
16. F. D'AURIA, M. LEONARDI, H. GLAESER, and R. POCHARD, "Current Status of Methodologies Evaluating the Uncertainty in the Prediction of Thermal-Hydraulic Phenomena in Nuclear Reactors," Proc. Two-Phase Flow Modeling and Experimentation 1995, Roma, Italy, October 9-11, 1995, p. 501, Edizioni ETS (1995).
17. A. PROâEK, M. KOZUH, J. MARN, B. MAVKO, S. PETELIN, and A. STRITAR, "Performance of ECCS During LOCA in a Two Loop PWR," Proc. Int. Conf. New Trends in Nuclear System Thermohydraulics, Pisa, Italy, May 30-June 2, 1994, Vol. 1, p. 509, Edizioni ETS (1994).
18. J. MARN, and R. P. PRIOR, "Determination Methodology of LOCA Break Sizes Classifying Small, Medium, and Large Breaks," Proc. Int. Conf. New Trends in Nuclear System Tliennohydraulics, Pisa, Italy, May 30-June 2, 1994, Vol. 1, p. 593, Edizioni ETS (1994).
19. F. D'AURIA, A. ERAMO, and W. GIANNOTTI, "Advancements in Planning of an IAU-Code," Proc. 4th Regional Mtg. Nuclear Energy in Central Europe, Bled, Slovenia, September 7-10,1997, p. 523, Nuclear Society of Slovenia (1997).
20. D. M. SNIDER, K. L. WAGNER, W. H. CRUSH, and K. R. JONES, "Nuclear Plant Analyzer," NUREG/CR-6291, INEL-94/0123, Vols. 1 through 4, Idaho National Engineering Laboratory (1995).
21. "Separate Effects Test Matrix for Thermal-Hydraulic Code Validation," OECD-NEA-CSNI R(93) 14, Organization for Economic Cooperation and Development, Nuclear Energy Agency, Committee on the Safety of Nuclear Installations (1993).
22. W. E. BURCHIL, "Physical Phenomena of a Small-Break Loss of Coolant Accident in a PWR," Nucl. Safety, 23, 525 (1982).
23. "Compendium of ECCS Research for Realistic LOCA Analysis," NUREG-1230, U.S. Nuclear Regulatory Commission (1988).
24. I. KLJENAK and A. PROSEK, "Analysis of Similarity Between Small-Break Loss-of-Coolant Accident Scenarios Based on Identification and Ranking of Basic Phenomena," Proc. 3rd Regional Mtg. Nuclear Energy in Central Europe, Portoroz, Slovenia, September 16-19, 1996, p. 285, Nuclear Society of Slovenia (1996).
25. N. AKSAN, F. D'AURIA, H. GLAESER, J. LILLINGTON, R. POCHARD, and A. SJOBERG, "Systematic Evaluation of the CSNI Separate Effects Tests Validation Matrix Data Base and Resulting Needs," Proc. 5th Int. Conf. Nuclear Engineering (ICONE-5), Nice, France, May 26-30,1997, American Society of Mechanical Engineers, French Nuclear Society, and Japan Society of Mechanical Engineers (1997).
26. I. KLJENAK and A. PROSEK, "Development of a Phenomena Identification and Ranking Table for a Small-Break Loss-of Coolant Accident Scenario," Proc. 4th Regional Mtg. Nuclear Energy in Central Europe, Bled, Slovenia, September 7-10, 1997, p. 515, Nuclear Society of Slovenia (1997).
27. RELAP5 CODE DEVELOPMENT TEAM, "RELAP5/ MOD3 Code Manual, Vols. 1 to 7," NUREG/CR-5535, U.S. Nuclear Regulatory Commission (1995).
28. "American National Standard for Decay Power in Light Water Reactors," ANSI/ANS-5.1-1979, American National Standards Institute/American Nuclear Society.
29. S. N. AKSAN, F. D'AURIA, and H. STÄDTKE, "User Effects on the Thermal-Hydraulic Transient System Code Calculations," J. Nucl. Eng. Des., 145, 159 (1993).
30. PARZER, A. PROSEK, M. SENCAR, A. STRITAR, B. SARLER, D. ZEBELJAN, and M. ZELJKO, "NPP Kräko Data Base for the Simulation with RELAP5/MOD2 Code," IJS-DP5713, Jozef Stefan Institute (1989) (in Slovene).
31. I. PARZER, S. PETELIN, and B. MAVKO, "Modelling Operator Rediagnosis During a SGTR Event," J. Nucl. Eng. Des., 159, 143 (1995).
32. S. PETELIN, B. MAVKO, and O. GORTNAR, "RELAP5/ MOD2 Split Reactor Vessel Model and Steamline Break Analysis," Nucl. Technol, 102, 116 (1993).
33. S.PETELIN,B.MAVKO,andl.TISELJ,"ParametricAnalysis of PWR NPP Feedwater System Pipe Break," Energia Nucleare, 10, 37 (1993).
34. B. MAVKO and A. PROSEK, "Peak Cladding Temperature Response Surface Generation Based on Simulations of a Small-Break Loss-of-Coolant Accident Scenario," Proc. 4th Regional Mtg. Nuclear Energy in Central Europe, Bled, Slovenia, September 7-10, 1997, p. 605, Nuclear Society of Slovenia (1997).
35. B. MAVKO and A. PROSEK, "Qualification of US Nodalization of Nuclear Power Plant at Steady-State Level," presented at Validation of Accident and Safety Analysis Methodology Mtg., Dubrovnik, Croatia, October 13-17, 1997.
36. T. VRABEC, "Failure of Main Steam Isolation Valve in Nuclear Power Plant," Diploma Thesis, University of Ljubljana (Nov. 1997) (in Slovene).
37. S. PETELIN, B. MAVKO, and O. GORTNAR, "Scaling for the Analysis of ISP-27 Accident Scenario Realized on Krsko NPP," Proc. Annual Mtg. Nuclear Society of Slovenia, Rogaska Slatina, Slovenia, September 18-21,1994, p. 269, Nuclear Society of Slovenia (1994).
38. S. PETELIN, B. MAVKO, O. GORTNAR, and I. RAVNIKAR, "BETHSY Experiment 9.1.b Parametric Analysis," Proc. Validation of Systems Transients Analysis Codes, Hilton Head, South Carolina, August 13-18, 1995, FED-Vol. 223, American Society of Mechanical Engineers (1995).
39. S. PETELIN, B. MAVKO, I. RAVNIKAR, P. CEBULL, and Y. A. HASSAN, "Nodalization Study for BETHSY Experiment," Proc. Int. Conf. Nuclear Engineering (ICONE-4), New Orleans, Louisiana, March 10-14,1996, Vol. 1, Part B, p. 955, American Society of Mechanical Engineers and Japan Society of Mechanical Engineers (1996).
40. I. RAVNIKAR, "Scaled Results Used on the Plant Model," MSc Thesis, University of Ljubljana (Jan. 1998) (in Slovene).
41. B. MAVKO, A. PROSEK, and F. D'AURIA, "Determination of Code Accuracy in Predicting Small-Break LOCA Experiment," Nucl. Technol, 120, 1 (1997).
42. F. D'AURIA, M. LEONARDI, and R. POCHARD, "Methodology for the Evaluation of Thermal-Hydraulic Codes Accuracy," Proc. Int. Conf. New Trends in Nuclear System Thermohydraulics, Pisa, Italy, May 30-June 2, 1994, p. 467, Edizioni ETS (1994).
43. Yankee Atomic Electric Company and Electricité de France, Personal Communication (1994).
44. A. PROSEK and B. MAVKO, "Evaluating Code Uncertainty-II: An Optimal Statistical Estimator Method to Evaluate the Uncertainties of Calculated Time Trends," Nucl. Technol, 126, 186 (1999).
45. M. HORVAT, M. LESKOVAR, and A. PROSEK, "Development, Description and Application of Code OSCPER," IJS-DP-7840, Jozef Stefan Institute (1998) (in Slovene).