Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
권호 표지
DOI QR코드

DOI QR Code

Uncertainties impact on the major FOMs for severe accidents in CANDU 6 nuclear power plant

  • R.M. Nistor-Vlad (Power Production and Usage Department, Faculty of Power Engineering, University Politehnica of Bucharest) ;
  • D. Dupleac (Power Production and Usage Department, Faculty of Power Engineering, University Politehnica of Bucharest) ;
  • G.L. Pavel (Power Production and Usage Department, Faculty of Power Engineering, University Politehnica of Bucharest)
  • Received : 2022.09.29
  • Accepted : 2023.04.12
  • Published : 2023.07.25
Download PDF
⟨ Previous Next ⟩

Abstract

In the nuclear safety studies, a new trend refers to the evaluation of uncertainties as a mandatory component of best-estimate safety analysis which is a modern and technically consistent approach being known as BEPU (Best Estimate Plus Uncertainty). The major objectives of this study consist in performing a study of uncertainties/sensitivities of the major analysis results for a generic CANDU 6 Nuclear Power Plant during Station Blackout (SBO) progression to understand and characterize the sources of uncertainties and their effects on the key figure-of-merits (FOMs) predictions in severe accidents (SA). The FOMs of interest are hydrogen mass generation and event timings such as the first fuel channel failure time, beginning of the core disassembly time, core collapse time and calandria vessel failure time. The outcomes of the study, will allow an improvement of capabilities and expertise to perform uncertainty and sensitivity analysis with severe accident codes for CANDU 6 Nuclear Power Plant.

Keywords

Acknowledgement

This research was carried by UPB as part of the IAEA Coordinated Research Project I31033 (Advancing the State-of-Practice in Uncertainty and Sensitivity Methodologies for Severe Accident Analysis in Water-Cooled Reactors) under the Agreement No 23710/R0.

References

  1. IAEA-TECDOC-1594, Analysis of Severe Accidents in Pressurized Heavy Water Reactors", Vienna, Austria, 2008, 978-92-0-105908-6, ISSN 1011-4289. 
  2. C.M. Allison, J.K. Hohorst, Role of RELAP/SCDAPSIM in Nuclear Safety, Science and Technology of Nuclear Installations, 2010, https://doi.org/10.1155/2010/425658, 2010, Article ID: 425658. 
  3. F. Zhou, D.R. Novog, L.J. Siefken, C.M. Allison, Development and benchmarking of mechanistic channel deformation models in RELAP/SCDAPSIM/MOD3.6 for CANDU severe accident analysis, Nucl. Sci. Eng. 190 (3) (2018) 209-237, https://doi.org/10.1080/00295639.2018.1442060. 
  4. The SCDAP/RELAP5 Development Team, SCDAP/RELAP5/MOD3.2 code manual. Volume II: damage progression model theory, ReVision 1 (1996). NUREG/CR-6150 INEL-96/0422. 
  5. IAEA-TECDOC-1727, Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications", Vienna, Austria, 2013, 978-92-0-114413-3, ISSN 1011-4289. 
  6. M. Mladin, D. Dupleac, I. Prisecaru, D. Mladin, Adapting and applying SCDAP/RELAP5 to CANDU in-vessel retention studies, Ann. Nucl. Energy 37 (6) (2010) 845-852, https://doi.org/10.1016/j.anucene.2010.02.015. 
  7. M. Perez-Ferragut, Integration of a Quantitative-Based Selection Procedure in an Uncertainty Analysis Methodology for NPP Safety Analysis", PhD Thesis, Nuclear Engineering Division, Politechnic University of Catalunya, 2012. 
  8. R.M. Nistor-Vlad, D. Dupleac, I. Prisecaru, M. Perez, C.M. Allison, J.K. Hohorst, CANDU 6 accident analysis using RELAP/SCDAPSIM with the integrated uncertainty package, in: Proceedings of the 26th International Conference on Nuclear Engineering, ICONE26, London, UK, 2018, https://doi.org/10.1115/ICONE26-82241. 
  9. Dupleac, D., Perez, M., Reventos, F., Allison, C. Uncertainty analysis of the 35% reactor inlet header break in a CANDU 6 reactor using RELAP/SCDAPSIM/MOD4.0 with integrated uncertainty analysis option. The 14th International Topical Meeting on Nuclear Reactor Thermalhydraulics (NURETH-14), Paper NURETH14-371. Canadian Nuclear Society, Toronto, Ontario (Canada). ISBN 978-1-926773-05-6. 
  10. Virgil E. Schrock, A revised ANS standard for decay heat from fission products, Nucl. Technol. 46 (2) (1979) 323-331, https://doi.org/10.13182/NT79-A32334. ISSN: 1943-7471. 
  11. T.S. Kwon, B.D. Chung, W.J. Lee, N.H. Lee, J.Y. Huh, Quantification of realistic discharge coefficient for the critical flow model of RELAP5/MOD3/KAERI, Journal of the Korean Nuclear Society 27 (1995). Available at: https://www.koreascience.or.kr/article/JAKO199511921630596.pdf. 
  12. R.S. Shewfelt, D.P. Godin, Verification Tests for GRAD, a Computer Program to Predict Nonuniform Deformation and Failure of Zr-wt2.5%Nb Pressure Tubes during a Postulated Loss of Coolant Accident, Atomic Energy of Canada Ltd (AECL), 1985. AECL-8384, https://inis.iaea.org/collection/NCLCollectionStore/_Public/16/078/16078232.pdf?r=1. 
  13. K. Pearson, Notes on Regression and Inheritance in the Case of Two Parents, Proceedings of the Royal Society of London, 1895, https://doi.org/10.1098/rspl.1895.0041. 
  14. Jerome L. Myers, Arnold D. Well, Research Design and Statistical Analysis, second ed., 2003, 978-0-8058-4037-7. 
  15. J.D. Evans, Straightforward Statistics for the Behavioral Sciences, CA Brooks/Cole Pub. Co. United States US, Pacific Grove, 1996, 9780534231002.