Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
권호 표지
DOI QR코드

DOI QR Code

Experimental approach to evaluate software reliability in hardware-software integrated environment

  • Received : 2019.07.21
  • Accepted : 2020.01.08
  • Published : 2020.07.25
Download PDF
⟨ Previous Next ⟩

Abstract

Reliability in safety-critical systems and equipment is of vital importance, so the probabilistic safety assessment (PSA) has been widely used for many years in the nuclear industry to address reliability in a quantitative manner. As many nuclear power plants (NPPs) become digitalized, evaluating the reliability of safety-critical software has become an emerging issue. Due to a lack of available methods, in many conventional PSA models only hardware reliability is addressed with the assumption that software reliability is perfect or very high compared to hardware reliability. This study focused on developing a new method of safety-critical software reliability quantification, derived from hardware-software integrated environment testing. Since the complexity of hardware and software interaction makes the possible number of test cases for exhaustive testing well beyond a practically achievable range, an importance-oriented testing method that assures the most efficient test coverage was developed. Application to the test of an actual NPP reactor protection system demonstrated the applicability of the developed method and provided insight into complex software-based system reliability.

Keywords

References

  1. T. Aldemir, D.W. Miller, M.P. Stovsky, J. Kirschenbaum, P. Bucci, A.W. Fentiman, L.T. Mangan, Current State of Reliability Modeling Methodologies for Digital Systems and Their Acceptance Criteria for Nuclear Power Plant Assessments" United States Nuclear Regulatory Commission, vol. 6901, U.S.NRC, USA, 2004. NUREG/CR.
  2. T. Aldemir, M.P. Stovsky, J. Kirschenbaum, D. Mandelli, P. Bucci, L.A. Mangan, D.W. Miller, X. Sun, E. Ekici, S. Guarro, M. Yau, B. Johnson, C. Elks, S.A. Arndt, Dynamic Reliability Modeling of Digital Instrumentation and Control Systems for Nuclear Reactor Probabilistic Risk Assessments, U.S.NRC, USA, 2004. NUREG/CR-6942.
  3. T. Aldemir, S. Guarro, J. Kirschenbaum, D. Mandelli, L.A. Mangan, P. Bucci, M. Yau, B. Johnson, C. Elks, E. Ekici, M.P. Stovsky, D.W. Miller, X. Sun, S.A. Amdt, Q. Nguyen, J. Dion, A Benchmark Implementation of Two Dynamic Methodologies for the Reliability Modeling of Digital Instrumentation and Control Systems, vol. 6985, U.S.NRC, , USA, 2004. NUREG/CR.
  4. T.L. Chu, M. Yue, G. Martinez-Guridi, K. Memick, J. Lehner, A. Kuritzky, Modeling a Digital Feedwater Control System Using Traditional Probabilistic Risk Assessment Methods, vol. 6997, U.S.NRC, , USA, 2004. NUREG/CR.
  5. S.J. Lee, W.D. Jung, J.E. Yang, PSA model with consideration of the effect of fault-tolerant techniques in digital I&C systems", Ann. Nucl. Energy 87 (2010).
  6. S.J. Lee, J.G. Choi, H.G. Kang, S.C. Jang, Reliability assessment method for NPP digital I&C systems considering the effect of automatic periodic tests, Ann. Nucl. Energy 37 (2010).
  7. H.G. Kang, M.C. Kim, S.J. Lee, H.J. Lee, H.S. Eom, J.G. Choi, S.C. Jang, An overview of risk quantification issues of digitalized nuclear power plants using static fault tree, Nucl. Eng. Technol. 41 (2009).
  8. H.G. Kang, T. Sung, An analysis of safety-critical digital systems for risk-informed design, Reliab. Eng. Syst. Saf. 78 (2002).
  9. J.H. Jo, S.J. Lee, W.D. Jeong, Fault analysis of reactor protection system based on FMEA, 2014. KAERI, ROK, KAERI/TR-5655.
  10. M.C. Kim, S.C. Jang, J. Ha, Possibilities and limitations of applying software reliability growth models to safety critical software, Nucl. Eng. Technol. 39 (2007).
  11. H.S. Eom, G.Y. Park, S.C. Jang, H.S. Son, H.G. Kang, V&V-based remaining fault estimation model for safety-critical software of a nuclear power plant, Ann. Nucl. Energy 51 (2013).
  12. T.L. Chu, Development of Quantitative Software Reliability Models for Digital Protection Systems of Nuclear Power Plants, NUREG/CR-7044, U.S. Nuclear Regulatory Commission, 2013.
  13. S. Kuball, J.H.R. May, A discussion of statistical testing on a safety-related application, Proc. Inst. Mech. Eng. O J. Risk Reliab. 221 (2007) 121-132.
  14. H.G. Kang, H.G. Lim, H.J. Lee, M.C. Kim, S.C. Jang, Input-profile-based software failure probability quantification for safety signal generation systems, Reliab. Eng. Syst. Saf. 94 (2009) 1542-1546. https://doi.org/10.1016/j.ress.2009.02.018
  15. S.H. Lee, S.J. Lee, J.K. Park, E.C. Lee, H.G. Kang, Development of simulation-based testing environment for safety-critical software, Nucl. Eng. Technol. 50 (2018).
  16. P.V. Varde, J.G. Choi, D.Y. Lee, J.B. Han, Reliability Analysis of Protection System of Advanced Pressurized Water Reactor - APR 1400 vol. 2468, Korea Atomic Energy Research Institute (KAERI), Republic of Korea (ROK), 2003. KAERI/TR.
  17. C.G. Lee, I.S. Oh, D.H. Kim, J.H. Park, J.H. Shin, Y.B. Kim, Requirements for the Development of KNICS Control Systems, 2004. KAERI, ROK, KAERI/TR-2737.
  18. J.B. Yoo, S.D. Cha, E.K. Jee, in: A Verification Framework for FBD Based Software in Nuclear Power Plants" 15th Asia-Pacific Software Engineering Conference, 2008.
  19. S.J. Lee, W. Jung, J. Yang, PSA Model Considering the Effects of Self-Diagnostic of Digital I&C Systems, Korea Atomic Energy Research Institute (KAERI), Republic of Korea (ROK), 2015. KAERI/TR-5946.