Journal of Nuclear Science, Engineering and Technology (JONSAT)
In cooperation with the Iranian Nuclear Society

A comparison of static and dynamic event tree analyses for SBO accident in VVER-1000/V446 NPP

Document Type : Research Paper

Authors

M.E. Amirsoltani
A. Pirouzmand
M.R. Nematollahi

Department of Nuclear Engineering, School of Mechanical Engineering, Shiraz University, P.O.BOX: 71946-84334, Shiraz - Iran

https://doi.org/10.24200/nst.2022.1455
Abstract
Event tree analysis is applied to quantify the core damage frequency (CDF) and assess the risk of nuclear power plants (NPPs) resulting from various postulated initiating events. To calculate this criterion, it is necessary to generate the probable scenarios according to the function of safety systems and the operator's actions. The classical event tree is currently used in PSA analysis. This method does not consider the accident's dynamics and scenarios. It considers only the availability/unavailability of the safety system functions and the operator's actions to calculate the frequency of each scenario. In contrast, the dynamic event tree method applies physical and probabilistic models to generate branches in the event tree, calculate the frequency of each scenario, determine the time profile of core damage, and time variation of physical parameters of the NPP for each scenario. This paper develops the dynamic event tree for the SBO accident at the VVER-1000/V446 NPP using the RELAP5 and RAVEN codes. The results are then compared with the outputs of the classical event tree. The results show that according to the assumptions, 3170 scenarios are evaluated in the dynamic event tree, while only 33 predetermined scenarios are examined in the conventional event tree. The calculated core damage frequencies are 3.61×10-6 (yr-1) and 1.97×10-6 (yr-1) for conventional and dynamic event trees, respectively.

Highlights

1. International Atomic Energy Agency (IAEA), Applications of probabilistic safety assessment (PSA) for nuclear power plants, IAEA-TECDOC-1200, VIENNA. (2001).

 

2. Burns III, Robert D., WASH 1400-reactor safety study, Prog. Nucl. Energy, 6 (1-3), 117-140 (1980).

 

3. R.J. Breeding, et al., The NUREG-1150 probabilistic risk assessment for the Surry Nuclear Power Station, Nuclear Engineering and Design, 135 (1), 29-59 (1992).

 

4. International Atomic Energy Agency, Development and Application of Level 1 Probabilistic Safety Assessment for Nuclear Power Plants Specific Safety Guide, International Atomic Energy Agency, (2010).

 

5. Nuclear Regulatory Commission (NRC), Severe accident risks: An assessment for five US nuclear power plants, (Nuclear Regulatory Commission, 1990).

 

6. Nuclear Regulatory Commission, Guidelines for the preparation of emergency operating procedures. Resolution of comments on NUREG-0799. No. NUREG--0899., Nuclear Regulatory Commission, (1982).

 

7. C. ACOSTA, N. SIU, Dynamic Event Tree Analysis Method (DETAM) for Accident Sequence Analysis. NUREG- CR-5608, U.S. Nuclear Regulatory Commission, (1992).

 

8. C. ACOSTA, N. SIU, Dynamic Event Tree in Accident Sequence Analysis: Application to Steam Generator Tube Rupture, Reliab. Eng. Syst. Safety, 41, 135 (1993).

 

9. N. SIU, Risk Assessment for Dynamic Systems: An Overview, Reliab. Eng. Syst. Safety, 43, 43 (1994).

 

10. C.C. Chao, C.J. Chang, Development of a dynamic event tree for a pressurized water reactor steam generator tube rupture event, Nuclear Technology, 130(1), 27-38 (2000).

 

11. Y.H.J. Chang, A. Mosleh, Cognitive Modeling and Dynamic Probabilistic Simulation of Operating Crew Response to Complex System Accidents (Ads-Idacrew), Center for Technology Risk Studies, University of Maryland, College Park, Maryland. (1999).

 

12. K.S. Hsueh, A. Mosleh, The development and application of the accident dynamic simulator (ADS) for dynamic probabilistic risk assessment of nuclear power plants, Reliability Engineering and System Safety, 52, 297 (1996).

 

13. M. Hakobyan, et al., Dynamic generation of accident progression event trees, Nuclear Engineering and Design, 12, 238 (2008).

 

14. Z.K. Jankovsky, et al., Extension of the ADAPT Framework for Multiple Simulators, (Sandia National Laboratory, United States, 2016).

 

15. Z.K. Jankovsky, et al, Dynamic Importance Measures in the ADAPT Framework, (Sandia National Laboratory, United States, 2016).

 

16. Z.K. Jankovsky, et al., Conditional Tree Reduction in the ADAPT Framework, (Sandia National Laboratory, United States, 2016).

 

17. T.A. Wheeler, et al., Nuclear Power Plant Cyber Security Discrete Dynamic Event Tree Analysis, (Sandia National Laboratory, United States, 2017).

 

18. V. Rychkov, et al., ADAPT-MAAP4 coupling for a dynamic event tree study, International Topical Meeting on Probabilistic Safety Assessment and Analysis, 1, 140 (2015).

 

19. A. Alfonsi, et al., Dynamic event tree analysis through Raven, Proceedings of ANS PSA 2013 International Topical Meeting on Probabilistic Safety Assessment and Analysis, (2013).

 

20. C. Picoco, et al., Coupling of RAVEN and MAAP5 for the dynamic event tree analysis of nuclear power plants, The Second International Conference on Engineering Sciences and Technologies, (2017).

 

21. A. Alfonsi, et al., RAVEN: Development of the adaptive dynamic event tree approach, Tech. Rep. INL/MIS-14-33246, Idaho National Laboratory (INL), (2014).

 

22. A. Alfonsi, et al., RAVEN: Dynamic Event Tree Approach Level III Milestone, No. INL/EXT-13-30332, Idaho National Laboratory (INL), (2013).

 

23. A. Alfonsi, et al., Hybrid Dynamic Event Tree sampling strategy in RAVEN code, Proceedings of International Topical Meeting on Probabilistic Safety Assessment, 26 (2015).

 

24. A. Alfonsi, Andrea, et al., RAVEN and dynamic probabilistic risk assessment: Software overview, Proceedings of ESREL European Safety and Reliability Conference, (2014).

 

25. A. Alfonsi, et al., Raven theory manual, No. INL/EXT-16-38178. Idaho National Lab. (INL), Idaho Falls, ID (United States), (2016).

 

26. A. Alfonsi, et al., RAVEN theory manual and user guide, Idaho National Laboratory, Idaho Falls (2017).

 

27. International Atomic Energy Agency, Design of Electrical Power Systems for Nuclear Power Plants, IAEA Safety Standards Series No. SSG-34, IAEA, Vienna (2016).

 

28. International Atomic Energy Agency (IAEA). Design Provisions for Station Blackout at Nuclear Power Plants, IAEA-TECDOC-1770, VIENNA. (2015).

 

29. International Atomic Energy Agency, Accident Management Programs for Nuclear Power Plants, IAEA Safety Standards Series No. SSG-54, IAEA, Vienna (2019).

 

30. International Atomic Energy Agency (IAEA), The final safety analysis report of BNPP's VVER-1000 reactor, Accident Analysis, Moscow. Chapter 15, (2015).

 

31. International Atomic Energy Agency (IAEA), Probabilistic safety assessment of BNPP’s VVER-1000 reactor, Ministry of Russian Federation of Atomic Energy (2014).

 

32. International Atomic Energy Agency, The Fukushima Daiichi Accident, IAEA, Vienna, (2015).

 

33. Z. Tabadar, et al., Probabilistic safety assessment of portable equipment applied in VVER-1000/V446 nuclear reactor during loss of ultimate heat sink accident for stress test program development, Progress in Nuclear Energy, 117, 103101 (2019).

 

34. A. Pirouzmand, et al., Safety analysis of LBLOCA in BDBA scenarios for the Bushehr's VVER-1000 nuclear power plant, Progress in Nuclear Energy, 88, 231 (2016).

 

35. Z. Tabadar, et al., Thermal-hydraulic modeling for deterministic safety analysis of portable equipment application in the VVER-1000 nuclear reactor during loss of ultimate heat sink accident using RELAP5/MOD3. 2 code, Annals of Nuclear Energy, 127, 53 (2019).

 

36. M. Jabbari, et al., The safety improvement of VVER-1000 NPP against SBO accident using portable air-cooled diesel generator, Annals of Nuclear Energy, 139, 107225 (2020).

 

37. A. Ghasemi, et al., Investigation of core meltdown phenomena and radioactive materials release in VVER‐1000/V446 nuclear reactor at severe accident condition due to LBLOCA along SBO, International Journal of Energy Research, 44.10, 8113 (2020).

 

38. D. Gertman, et al., The SPAR-H human reliability analysis method, US Nuclear Regulatory Commission, 230, 35 (2005).

 

39. S. Rodgers, et al., Integrated power recovery using Markov modeling, Journal of Engineering for Gas Turbines and Power, 133, 11 (2011).

Keywords

Static Event Tree (ET)
Dynamic Event Tree (DET)
Station Black-Out (SBO) accident
VVER-1000/V446 NPP
RELAP5/mod3.2
Risk Analysis Virtual Environment (RAVEN)
1. International Atomic Energy Agency (IAEA), Applications of probabilistic safety assessment (PSA) for nuclear power plants, IAEA-TECDOC-1200, VIENNA. (2001).
 
2. Burns III, Robert D., WASH 1400-reactor safety study, Prog. Nucl. Energy, 6 (1-3), 117-140 (1980).
 
3. R.J. Breeding, et al., The NUREG-1150 probabilistic risk assessment for the Surry Nuclear Power Station, Nuclear Engineering and Design, 135 (1), 29-59 (1992).
 
4. International Atomic Energy Agency, Development and Application of Level 1 Probabilistic Safety Assessment for Nuclear Power Plants Specific Safety Guide, International Atomic Energy Agency, (2010).
 
5. Nuclear Regulatory Commission (NRC), Severe accident risks: An assessment for five US nuclear power plants, (Nuclear Regulatory Commission, 1990).
 
6. Nuclear Regulatory Commission, Guidelines for the preparation of emergency operating procedures. Resolution of comments on NUREG-0799. No. NUREG--0899., Nuclear Regulatory Commission, (1982).
 
7. C. ACOSTA, N. SIU, Dynamic Event Tree Analysis Method (DETAM) for Accident Sequence Analysis. NUREG- CR-5608, U.S. Nuclear Regulatory Commission, (1992).
 
8. C. ACOSTA, N. SIU, Dynamic Event Tree in Accident Sequence Analysis: Application to Steam Generator Tube Rupture, Reliab. Eng. Syst. Safety, 41, 135 (1993).
 
9. N. SIU, Risk Assessment for Dynamic Systems: An Overview, Reliab. Eng. Syst. Safety, 43, 43 (1994).
 
10. C.C. Chao, C.J. Chang, Development of a dynamic event tree for a pressurized water reactor steam generator tube rupture event, Nuclear Technology, 130(1), 27-38 (2000).
 
11. Y.H.J. Chang, A. Mosleh, Cognitive Modeling and Dynamic Probabilistic Simulation of Operating Crew Response to Complex System Accidents (Ads-Idacrew), Center for Technology Risk Studies, University of Maryland, College Park, Maryland. (1999).
 
12. K.S. Hsueh, A. Mosleh, The development and application of the accident dynamic simulator (ADS) for dynamic probabilistic risk assessment of nuclear power plants, Reliability Engineering and System Safety, 52, 297 (1996).
 
13. M. Hakobyan, et al., Dynamic generation of accident progression event trees, Nuclear Engineering and Design, 12, 238 (2008).
 
14. Z.K. Jankovsky, et al., Extension of the ADAPT Framework for Multiple Simulators, (Sandia National Laboratory, United States, 2016).
 
15. Z.K. Jankovsky, et al, Dynamic Importance Measures in the ADAPT Framework, (Sandia National Laboratory, United States, 2016).
 
16. Z.K. Jankovsky, et al., Conditional Tree Reduction in the ADAPT Framework, (Sandia National Laboratory, United States, 2016).
 
17. T.A. Wheeler, et al., Nuclear Power Plant Cyber Security Discrete Dynamic Event Tree Analysis, (Sandia National Laboratory, United States, 2017).
 
18. V. Rychkov, et al., ADAPT-MAAP4 coupling for a dynamic event tree study, International Topical Meeting on Probabilistic Safety Assessment and Analysis, 1, 140 (2015).
 
19. A. Alfonsi, et al., Dynamic event tree analysis through Raven, Proceedings of ANS PSA 2013 International Topical Meeting on Probabilistic Safety Assessment and Analysis, (2013).
 
20. C. Picoco, et al., Coupling of RAVEN and MAAP5 for the dynamic event tree analysis of nuclear power plants, The Second International Conference on Engineering Sciences and Technologies, (2017).
 
21. A. Alfonsi, et al., RAVEN: Development of the adaptive dynamic event tree approach, Tech. Rep. INL/MIS-14-33246, Idaho National Laboratory (INL), (2014).
 
22. A. Alfonsi, et al., RAVEN: Dynamic Event Tree Approach Level III Milestone, No. INL/EXT-13-30332, Idaho National Laboratory (INL), (2013).
 
23. A. Alfonsi, et al., Hybrid Dynamic Event Tree sampling strategy in RAVEN code, Proceedings of International Topical Meeting on Probabilistic Safety Assessment, 26 (2015).
 
24. A. Alfonsi, Andrea, et al., RAVEN and dynamic probabilistic risk assessment: Software overview, Proceedings of ESREL European Safety and Reliability Conference, (2014).
 
25. A. Alfonsi, et al., Raven theory manual, No. INL/EXT-16-38178. Idaho National Lab. (INL), Idaho Falls, ID (United States), (2016).
 
26. A. Alfonsi, et al., RAVEN theory manual and user guide, Idaho National Laboratory, Idaho Falls (2017).
 
27. International Atomic Energy Agency, Design of Electrical Power Systems for Nuclear Power Plants, IAEA Safety Standards Series No. SSG-34, IAEA, Vienna (2016).
 
28. International Atomic Energy Agency (IAEA). Design Provisions for Station Blackout at Nuclear Power Plants, IAEA-TECDOC-1770, VIENNA. (2015).
 
29. International Atomic Energy Agency, Accident Management Programs for Nuclear Power Plants, IAEA Safety Standards Series No. SSG-54, IAEA, Vienna (2019).
 
30. International Atomic Energy Agency (IAEA), The final safety analysis report of BNPP's VVER-1000 reactor, Accident Analysis, Moscow. Chapter 15, (2015).
 
31. International Atomic Energy Agency (IAEA), Probabilistic safety assessment of BNPP’s VVER-1000 reactor, Ministry of Russian Federation of Atomic Energy (2014).
 
32. International Atomic Energy Agency, The Fukushima Daiichi Accident, IAEA, Vienna, (2015).
 
33. Z. Tabadar, et al., Probabilistic safety assessment of portable equipment applied in VVER-1000/V446 nuclear reactor during loss of ultimate heat sink accident for stress test program development, Progress in Nuclear Energy, 117, 103101 (2019).
 
34. A. Pirouzmand, et al., Safety analysis of LBLOCA in BDBA scenarios for the Bushehr's VVER-1000 nuclear power plant, Progress in Nuclear Energy, 88, 231 (2016).
 
35. Z. Tabadar, et al., Thermal-hydraulic modeling for deterministic safety analysis of portable equipment application in the VVER-1000 nuclear reactor during loss of ultimate heat sink accident using RELAP5/MOD3. 2 code, Annals of Nuclear Energy, 127, 53 (2019).
 
36. M. Jabbari, et al., The safety improvement of VVER-1000 NPP against SBO accident using portable air-cooled diesel generator, Annals of Nuclear Energy, 139, 107225 (2020).
 
37. A. Ghasemi, et al., Investigation of core meltdown phenomena and radioactive materials release in VVER‐1000/V446 nuclear reactor at severe accident condition due to LBLOCA along SBO, International Journal of Energy Research, 44.10, 8113 (2020).
 
38. D. Gertman, et al., The SPAR-H human reliability analysis method, US Nuclear Regulatory Commission, 230, 35 (2005).
 
39. S. Rodgers, et al., Integrated power recovery using Markov modeling, Journal of Engineering for Gas Turbines and Power, 133, 11 (2011).
Journal of Nuclear Science, Engineering and Technology (JONSAT)
Volume 43, Issue 3 - Serial Number 101
October 2022
Pages 133-146

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