RELAP5 simulation of surge line break accident using combined and best estimate plus uncertainty approaches

Nuclear Engineering and Design

Volumn 239, Issue 11, 2009, Pages 2500-2513

  Krištof, Marián ^a,d   Kliment, Tomáš ^b   Petruzzi, Alessandro ^c   Lipka, Jozef ^d  

^a AiNS   (Slovakia)

^b VUJE   (Slovakia)

^c UNIVERSITY OF PISA   (Italy)

^d SLOVAK UNIVERSITY OF TECHNOLOGY   (Slovakia)

Author keywords

[No Author keywords available]

Indexed keywords

BEST ESTIMATES; COMPUTER CODES; EXPERIMENTAL VALUES; HARDWARE DIFFERENCES; INTEGRAL TEST; PEAK CLADDING TEMPERATURE; PLANT MODEL; POWER-SCALING; QUALITATIVE ASSESSMENTS; SAFETY ANALYSIS; SAFETY MARGIN; SIMULATION-BASED; UNCERTAINTY EVALUATION;

SIMULATORS; TEST FACILITIES;

UNCERTAINTY ANALYSIS;

EID: 70349316267     PISSN: 00295493     EISSN: None     Source Type: Journal    
DOI: 10.1016/j.nucengdes.2009.07.003     Document Type: Article

References (23)

1. Bovalini R., and D'Auria F. Scaling of the accuracy of RELAP5/mod2 Code. J. Nucl. Eng. Des. 139 2 (1993) 187-204
2. Boyack B.E., et al. Quantifying reactor safety margins. J. Nucl. Eng. Des. 119 (1990) 1-15
3. D'Auria F., et al. Overview of uncertainty issues and methodologies. OECD/CSNI Seminar on Best Estimate Methods in Thermal Hydraulic Safety Analyses. Ankara (1998)
4. D'Auria F., and Galassi G.M. Code validation and uncertainties in system thermal-hydraulics. J. Prog. Nucl. Energy 33 1/2 (1998) 175-216
5. D'Auria F., and Giannotti W. Development of code with capability of internal assessment of uncertainty. J. Nucl. Technol. 131 1 (2000) 159-196
6. Glaeser H. Description and Application of GRS Method for Uncertainty and Sensitivity Evaluation of Code Results (2005), IAEA RW on Application of Deterministic Safety Analysis, Budapest
7. Husárček J., and Krištof M. Requirements on Deterministic Safety Analysis of Nuclear Power Plants (2008), ÚJD SR Regulatory Guide BNS I.11.1/2008
8. International Atomic Energy Agency, 2001. Safety Assessment and Verification for Nuclear Power Plants. Safety Standard Series No. NS-G-1.2, 2001.
9. International Atomic Energy Agency, 2008. Best Estimate Safety Analysis for Nuclear Power Plants: Uncertainty Evaluation. IAEA Safety Series Report No. 52, Vienna.
10. Kimber, G.R., 1998. Phare 4.2.6b: VVER-440/213 Accident Analysis DBA Code Verification. AEA Technology Report.
11. Krištof M., and Vojtek I. Uncertainty analyses for the large break LOCA in VVER-440/213. EUROSAFE Forum 2003, Proceedings, vol. II. Paris (2003) pp.45
12. Krištof M. Application of BEPU methodology for the licensing purposes in Slovakia. BE2004 - International Meeting on Updates in Best Estimate Methods in Nuclear Installation Safety Analysis, Proceedings. Washington, DC (2004) pp.124
13. Krištof, M., 2006. Uncertainty analysis of LOFT L2-3 experiment with the GRS uncertainty Method. Technical Report No. 240-07/2006, ÚJD SR.
14. Krištof, M., Petruzzi, A., 2006. Uncertainty analysis of LOFT L2-3 experiment with the CIAU uncertainty method. Technical Report No. 240-02/2006, ÚJD SR.
15. Krištof M. Best Estimate Safety Analysis with Evaluation of Uncertainty (2007), ÚJD SR Regulatory Guide BNS I.11.3/2007.
16. NEA, 1997. Proceedings of OECD/NEA/CSNI Workshop. Transient Thermal-Hydraulic and Neutronics Codes Requirements. Annapolis, November 5-8, 1996, NUREG/CP-0159, NEA/CSNI/R(97) 4.
17. Petruzzi A., et al. Methodology of internal assessment of uncertainty and extension to neutron-kinetics/thermal-hydraulics coupled codes. Nucl. Sci. Eng. (2005) 149
18. Petruzzi A., et al. International training program: 3D S.UN.COP - scaling, uncertainty and 3D thermal-hydraulics/neutron-kinetics coupled codes seminars. 6th International Conference on Nuclear Option in Countries with Small and Medium Electricity Grids, Dubrovnik; ANS 2006 Annual Meeting, Reno, June 4-8, 2006 and ICONE-14, Miami, July 17-20 (2006)
19. Petruzzi A., and D'Auria F. Thermal-Hydraulic System Codes in Nuclear Reactor Safety and Qualification Procedures (2008), Hindawi Publishing Corporation Science and Technology of Nuclear Installations Volume 2008 Article ID 460795
20. Prošek A., et al. Quantitative Assessment of MCP Trip Transient in a VVER. J. Nucl. Eng. Des. 227 (2004) 85-96
21. Safety Analysis Report, 2007. Mochovce NPP Unit 1 and 2. Mochovce NPP.
22. Szabados L., et al. Results of the Experiments Performed in the PMK-2 Facility for VVER Safety Studies (2007), Akadémiai Kiadó, Budapest
23. US Nuclear Regulatory Commissions, 10 CFR 50, 1996. Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors and Appendix K, "ECCS Evaluation Models", to 10 CFR Part 50. Code of Federal Regulations.