DOI QR코드

DOI QR Code

Structural safety reliability of concrete buildings of HTR-PM in accidental double-ended break of hot gas ducts

  • Received : 2018.11.21
  • Accepted : 2019.10.21
  • Published : 2020.05.25

Abstract

Safety analysis of nuclear power plant (NPP) especially in accident conditions is a basic and necessary issue for applications and commercialization of reactors. Many previous researches and development works have been conducted. However, most achievements focused on the safety reliability of primary pressure system vessels. Few literatures studied the structural safety of huge concrete structures surrounding primary pressure system, especially for the fourth generation NPP which allows existing of through cracks. In this paper, structural safety reliability of concrete structures of HTR-PM in accidental double-ended break of hot gas ducts was studied by Exceedance Probability Method. It was calculated by Monte Carlo approaches applying numerical simulations by Abaqus. Damage parameters were proposed and used to define the property of concrete, which can perfectly describe the crack state of concrete structures. Calculation results indicated that functional failure determined by deterministic safety analysis was decided by the crack resistance capability of containment buildings, whereas the bearing capacity of concrete structures possess a high safety margin. The failure probability of concrete structures during an accident of double-ended break of hot gas ducts will be 31.18%. Adding the consideration the contingency occurrence probability of the accident, probability of functional failure is sufficiently low.

References

  1. Z. Wu, D. Lin, D. Zhong, The design features of the HTR-10, Nucl. Eng. Des. 218 (2002) 25-32.
  2. Z. Zhang, Z. Wu, D.Wang, Y. Xu, Y. Sun, F. Li, et al., Current status and technical description of Chinese 2 $\times$ 250 MW th HTR-PM demonstration plant, Nucl. Eng. Des. 239 (2009) 1212-1219.
  3. C. Fang, X. Bao, C. Yang, Y. Yang, J. Cao, The R&D of HTGR high temperature helium sampling loop: from HTR-10 to HTR-PM, Nucl. Eng. Des. 306 (2016) 192-197.
  4. A. Dvorzhak, J.C. Mora, B. Robles, C.G. Soares, Probabilistic risk assessment from potential exposures to the public applied for innovative nuclear installations, Reliab. Eng. Syst. Saf. 152 (2016) 176-186.
  5. R. Barati, S. Setayeshi, Functional reliability evaluation of an MTR-pool type research reactor core using the loadecapacity interference model, Ann. Nucl. Energy 58 (2013) 151-160.
  6. L.P. Pagani, G.E. Apostolakis, P. Hejzlar, The impact of uncertainties on the performance of passive systems, Nucl. Technol. 149 (2005) 129-140.
  7. W. Peng, T. Chen, Q. Sun, J. Wang, S. Yu, Preliminary experiment design of graphite dust emission measurement under accident conditions for HTGR, Nucl. Eng. Des. 316 (2017) 218-227.
  8. A.A. Ryabov, V.I. Romanov, S.S. Kukanov, K.V. Tsiberev, S.V. Evropin, A.P. Rakhmanov, Numerical study of the dynamic strength of RBMK-1000 stack with fuel channel rupture, Atom. Energy (2018) 1-6.
  9. Z. Gao, Z. Jiang, B. Li, C. Wang, Safety analysis of accident scenarios for the HTR-10, Tsinghua Sci. Technol. 1 (1996) 27-31.
  10. G. Wang, S. Niu, R. Cao, Summary of severe accident issues of LBE-cooled reactors, Ann. Nucl. Energy (2018) 531-539.
  11. X. Li, L.I. Shi, Z. Zhang, S. He, Safety analysis for hot gas duct vessel in HTR-PM, Nucl. Technol. 174 (2011) 29-40.
  12. S. Zhang, J. Tong, J. Zhao, An integrated modeling approach for event sequence development in multi-unit probabilistic risk assessment, Reliab. Eng. Syst. Saf. 155 (2016) 147-159.
  13. Shinozuka M, Kako T, Hwang H, Reich M. Development of a Reliability-Analysis Method for Category I Structures. International Conference on Structural Mechanics in Reactor Technology. United States 1983. p. 1-8.
  14. F. Sanchez-Saez, A.I. Sꠑanchez, J.F. Villanueva, S. Carlos, S. Martorell, Uncertainty analysis of a large break Loss of coolant accident in a pressurized water reactor using non-parametric methods, Reliab. Eng. Syst. Saf. 174 (2018) 19-28.
  15. F.S. D'Auria, H. Glaeser, S. Lee, J. Miak, M. Modro, R. Schultz, Best Estimate Safety Analysis for Nuclear Power Plants: Uncertainty Evaluation, IAEA Safety Report Series: IAEA, 2008.
  16. X. Zheng, H. Itoh, H. Tamaki, M. Yu, An integrated approach to source term uncertainty and sensitivity analyses for nuclear reactor severe accidents, J. Nucl. Sci. Technol. 53 (2016) 333-344.
  17. J.F.E. Briesmeister, MCNP - A General Monte Carlo N-Particle Transport Code, 2003.
  18. W. Wang, F.D. Maio, E. Zio, Three-Loop Monte Carlo simulation approach to multi-state physics modeling for system reliability assessment, Reliab. Eng. Syst. Saf. 167 (2017) 276-289.
  19. Y. Fukano, SAS4A analysis on hypothetical total instantaneous flow blockage in SFRs based on in-pile experiments, Ann. Nucl. Energy 77 (2015) 376-392.
  20. B.S. Anis, J. Areeya, H. Tewfik, J.R. White, Assessment of RELAP5 model for the university of Massachusetts lowell research reactor, Nucl. Technol. Radiat. Prot. 21 (2006) 3-12.
  21. R. Li, X.N. Chen, L. Andriolo, A. Rineiski, 3D numerical study of LBE-cooled fuel assembly in MYRRHA using SIMMER-IV code, Ann. Nucl. Energy 104 (2017) 42-52.
  22. H.P. Lee, Shell finite element of reinforced concrete for internal pressure analysis of nuclear containment building, Nucl. Eng. Des. 241 (2011) 515-525.
  23. C. Zhao, J. Chen, Numerical simulation and investigation of the base isolated NPPC building under three-directional seismic loading, Nucl. Eng. Des. 265 (2013) 484-496.
  24. J. Lubliner, J. Oliver, S. Oller, E. Onate, A plastic-damage model for concrete, Int. J. Solids Struct. 25 (1989) 299-326.
  25. J. Lee, G.L. Fenves, Plastic-damage model for cyclic loading of concrete structures, J. Eng. Mech. 124 (1998) 892-900.
  26. GB50010. Code for Design of Concrete Structures, China Architecture & Building Press, Beijing, 2010 (in Chinese).