Transactions of the Korean Society of Mechanical Engineers A (대한기계학회논문집A)

The Korean Society of Mechanical Engineers (대한기계학회)
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Development of Lifetime Evaluation and Management Technologies for Nuclear Power Plants

원자력발전소 수명평가 및 수명관리 기술개발

  • 진태은 (한국전력기술(주) 기계기술처)
  • Published : 2009.10.01
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Abstract

Operating experience of the various components in the nuclear power plants has shown that a variety of degradation mechanisms can occur during operation. Therefore, the accurate lifetime evaluation and systematic management are very important for the safe as well as the economical operation of the nuclear power plants. In this paper, the characteristics of a total of 17 degradation mechanisms were reviewed and the plausible degradation mechanisms such as stress corrosion cracking, fatigue, irradiation embrittlement, and so on, were identified. Also, the lifetime evaluation technologies which have been developed for the application to the domestic nuclear power plants are described. In addition, a total of 48 aging management programs which have been established for the safe operation of the various components are explained.

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References

  1. 10 CFR Part 54, 1995, 'Nuclear Plant License Renewal,' Federal Register, Vol. 60, No. 88
  2. Atomic Energy Act, 2008, 'Periodic Safety Review,' Article 23-3
  3. MEST, 2008, 'Guidelines for the Application of Codes and Standards in the Evaluation of the Continued Operation for Nuclear Facilities,' Notice 2008-17
  4. ASME, 2007, 'Rules for Construction of Nuclear Power Plant Components,' Sec. III, Division I
  5. ASME, 2007, 'Environmental Effects on Components,' Sec. III, App. W
  6. Jin, T.E., et. al. 2009, 'Current Status and Prospect for Periodic Safety Review of Aging Nuclear Oiwer Plants in Korea,' Nuclear Engineering and Technology, Vol. 41, No. 4, pp 545-548 https://doi.org/10.5516/NET.2009.41.4.545
  7. Chung, H.D., 2009, 'Regulatory Policies on the Aging Managements for the Nuclear Components,' 16th Symposium on the Nuclear Components Integrity, pp. 17-30
  8. Park, Y.K., 2009, 'Integrated Management Plan on the Aging Managements of the Metallic Materials in NPPs,' 16th Symposium on the Nuclear Components Integrity, pp. 31-44
  9. USNRC, 2005, 'Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plant,' NUREG-1800, Rev. 1
  10. ASME, 2006, 'Alternative Examination Requirement for PWR Reactor Vessel Upper Heads with Nozzle Having Pressure Retaining Partial Penetration Weld,' Code Case N-729-1
  11. USNRC, 1988, 'Thermal Stresses in Piping Connected to Reactor Coolant Systems,' Bulletin 88-08
  12. EPRI, 2000, 'MRP-25: Operating Experience Regarding Thermal Fatigue of Unisolable Piping Connected to PWR Reactor Coolant Systems,' TR-1001006
  13. ASME, 2007, 'Design Stress Intensity Values, Allowable Stresses, Material Properties, and Design Fatigue Curves,' Sec. III, App. I
  14. USNRC, 2007, 'Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials,' NUREG/CR-6909
  15. Higuchi, M., Nakamura, T. and Sugie, Y., 2009, 'Updated Knowledge Implemented to the Revision of Environmental Fatigue Evaluation Method for Nuclear Power Plants in JSME Code,' ASME PVP Conference, No. 77077
  16. USNRC, 2007, 'Guidelines for Evaluating Fatigue Analyses Incorporating the Life Reduction of Metal Components due to the Effects of the Light-Water Reactor Environment for New Reactors,' Regulatory Guide 1.207
  17. USNRC, 1999, 'Effect of LWR Coolant Environments on Fatigue Design Curves of Austenitic Stainless Steels,' NUREG/CR-5704
  18. USNRC, 1997, 'Effect of LWR Coolant Environments on Fatigue Design Curves of Carbon and Low-Alloy Steels,' NUREG/CR-6583
  19. EPRI, 2001, 'User's Manual for Fatigue-Pro Version 3.0,' Report No. 1002861
  20. ASME, 2007, 'Assessment of Reactor Pressure Vessels with Low Upper Shelf Charpy Impact Energy Levels,' App. K
  21. USNRC, 1995, 'Evaluation of Reactor Pressure Vessel with Charpy Upper Shelf Energy Less than 50ft-lb,' Reg. Guide 1.161
  22. USNRC, 1987, 'Format of Content of Plant Specific PTS Safety Analysis Report for PWRs,' Reg. Guide 1.154
  23. Kim, H.S, Jung, S.G. and Jin, T.E., 2000, 'PREVIAS: Probabilistic Reactor Vessel Integrity Assessment System User's Manual,' KOPEC/ TR-2000-01
  24. USNRC, 1991, 'High Energy Piping Failures Caused by Wall Thinning,' IN 91-18
  25. EPRI, 1998, 'CHECWORKS Application Manager,' Version 1.0D, TR-103198-P3
  26. USNRC, 1994, 'Tensile Property Characteriz- ation of Thermally Aged Cast Stainless Steels,' NUREG/CR-6142
  27. USNRC, 1994, 'Estimation of Fracture Toughness of Cast Stainless Steels during Thermal Aging in LWR Systems,' NUREG/CR-6177
  28. USNRC, 2005, 'Generic Aging Lessons Learned (GALL) Report,' NUREG-1801, Rev. 1
  29. ASME, 2007, 'Evaluation of Flaws in Piping,' Sec. XI, App. C
  30. EPRI, 1997, 'Evaluation of Thermal Aging Embrittlement for Cast Austentitic Stainless Steel Components in LWR Reactor Coolant Systems,' TR-106092