Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
권호 표지

Treatment of Stainless Steel Cladding in Pressurized Thermal Shock Evaluation: Deterministic Analyses

  • Published : 2001.04.01
Download PDF
⟨ Previous Next ⟩

Abstract

Fracture mechanics is one of the major areas of the pressurized thermal shock (PTS) evaluation. To evaluate the reactor pressure vessel integrity associated with PTS, PFM methodology demands precise calculation of temperature, stress, and stress intensity factor for the variety of PTS transients. However, the existence of stainless steel cladding, with different thermal, physical, and mechanical property, at the inner surface of reactor pressure vessel complicates the fracture mechanics analysis. In this paper, treatment schemes to evaluate stress and resulting stress intensity factor for RPV with stainless steel clad are introduced. For a reference transient, the effects of clad thermal conductivity and thermal expansion coefficients on deterministic fracture mechanics analysis are examined.

Keywords

References

  1. USNRC, 'NRC Staff Evaluation of Pressurized Thermal Shock,' SECY 82-465(1982)
  2. R. D. Cheverton and D. G. Ball, 'OCA-P, A Deterministic and Probabilistic Fracture-Mechanics Code for Application to Pressure Vessels,' NUREG/CR-3618,OCA-P(1984)
  3. F. A. Simonen et al. 'VISA-Ⅱ, A Computer Code for Predicting the Probability of Reactor Vessel Failure,' NUREG/CR-4486(1986)
  4. T. L. Dickson, 'FAVOR: A Fracture Analysis Code for Nuclear Reactor Pressure Vessels, Release,' ORNC/NRC/LTR/94/1(1994)
  5. X. R. Wu and A. J. Carlsson, 'Weight Functions and Stress Intensity Factor Solutions,' Pergamon Press, New York(1991)
  6. T. Fett and D. Munz, 'Stress Intensity Factors and Weight Functions,' Computational Mechanics Publication, Boston(1997)
  7. C. H. Jang, et. al, Comparative Study of Probabilistic Fracture Mechanics Codes, in the Proceedings of the KNS '99 spring meeting, Pohang, Korea, May 28-29(1999)
  8. S. P. Timoshenko and J.N. Goodier, 'Theory of Elasticity,' McGraw-Hill Company, New York(1970)
  9. C. H. Jang, 'Treatment of Cladding on Pressurized Thermal Shock Evaluation,' KEPRITM.96NJ12.P1998.898(1998)
  10. C. H. Jang, 'Development of Probabilistic Fracture Mechanics Analysis Code: KAPTSKEPRI Analysis of Pressurized Thermal Shock,' KEPRI TM.97NJ26.P1999.511(1999)
  11. ASME, 'Analysis of Flows,' Nonmandatory Appendix A to ASME Boiler and Pressure Vessel Code, Section XI(1998)
  12. J. J. McGowan, Application of Warm Prestressing Effects to Fracture Mechanics Analysis of Nuclear Reator Vessels during Severe Thermal Shock, Nuclear Engineering and Design, Vol. 51,, pp. 431-444(1979) https://doi.org/10.1016/0029-5493(79)90131-6
  13. D. A. Curry, A model for Predicting the Influence of Warm Pre-stressing and Strain Aging on the Cleavage Fractrue Touhness of Ferritic Steels, International Journal of Fracture, Vol. 22, pp. 145-159(1983) https://doi.org/10.1007/BF00942720
  14. M. J. Jhung, Y. W. Park, and C. H. Jang, Pressurized Thermal Shock Analysis of A Reactor Pressure Vessel Using Critical Crack Depth Diagram, Int. J. of Pressure Vessels and Pipings, Vol. 76, pp. 813-823(1999) https://doi.org/10.1016/S0308-0161(99)00063-0
  15. K. Balkey et al., 'Documentation of Probabilistic Fracture Mechanics Codes Used for Reactor Pressure Vessels Subjected to Pressurized Thermal Shock Loading,' EPRI TR-105001(1995)