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Assessment of Steam Generator Tubes with Multiple Axial Through-Wall Cracks

축방향 다중관통균열이 존재하는 증기발생기 세관 평가법

  • 문성인 (성균관대학교 기계공학부) ;
  • 장윤석 (성균관대학교 기계공학부) ;
  • 김영진 (성균관대학교 기계공학부) ;
  • 이진호 (한국원자력안전기술원) ;
  • 송명호 (한국원자력안전기술원) ;
  • 최영환 (한국원자력안전기술원)
  • Published : 2004.11.01

Abstract

It is commonly requested that the steam generator tubes wall-thinned in excess of 40% should be plugged. However, the plugging criterion is known to be too conservative for some locations and types of defects and its application is limited to a single crack in spite of the fact that the occurrence of multiple through-wall cracks is more common in general. The objective of this research is to propose the optimum failure prediction models for two adjacent through-wall cracks in steam generator tubes. The conservatism of the present plugging criteria was reviewed using the existing failure prediction models for a single crack, and six new failure prediction models for multiple through-wall cracks have been introduced. Then, in order to determine the optimum ones among these new local or global failure prediction models, a series of plastic collapse tests and corresponding finite element analyses for two adjacent through-wall cracks in thin plate were carried out. Thereby, the reaction force model, plastic zone contact model and COD (Crack-Opening Displacement) base model were selected as the optimum ones for assessment of steam generator tubes with multiple through-wall cracks. The selected optimum failure prediction models, finally, were used to estimate the coalescence pressure of two adjacent through-wall cracks in steam generator tubes.

Keywords

Steam Generator Tube;Plastic Collapse;Failure Prediction Model;Limit Load Method;Plugging Criteria;Tube Rupture;Interaction Effect

References

  1. USNRC, 1996, 'Steam Generator Tube Failures,' NUREG/CR6365
  2. USNRC, 1976, 'Bases for Plugging Degraded PWR Steam Generator Tubes,' Regulatory Guide 1.121
  3. ASME, 1998, 'Rules for Construction of Nuclear Power Plant Components,' ASME Boiler and Pressure Vessel Code, Section III
  4. Cochet, B. and Flesch, B., 1987, 'Crack Stability Criteria in Steam Generator Tubes,' 9th Int. Conference on SMiRT, Vol. D, pp. 413-419
  5. Yu, Y.J., Kim, J.H., Kim, Y. and Kim, Y.J., 1994, 'Development of Steam Generator Tube Plugging Criteria for Axial Crack,' ASME PVP, Vol. 280, pp. 79-83
  6. Kim, H.D., Chung, H.S. and Hong, S.R., 1999, 'Discussion on Operation Leakage Criteria of Ulchin Unit 1&2 Steam Generators,' Proceedings of the Korean Nuclear Society Autumn Meeting
  7. Kim, H.D, Kim, K.T. and Chung, H.S, 1999, 'Structural Integrity Assessment on Axial PWSCC of Steam Generator Tubes,' Proceedings of the Korean Nuclear Society Autumn Meeting
  8. Gorman, J.A., Harris, J.E. and Lowenstein, D.B., 1995, 'Steam Generator Tube Fitness-for-Service Guidelines,' AECB Report, No. 2.228.2
  9. Lee, J.H., Park, Y.W., Song, M.H., Kim, Y.J. and Moon, S.I., 2000, 'Determination of Equivalent Single Crack Based on Coalescence Criterion of Collinear Axial Cracks,' Nuclear Engineering and Design, Vol. 205, pp. 1-11 https://doi.org/10.1016/S0029-5493(00)00368-X
  10. Kim, J.S. et al., 1999, 'Investigation Report for Steam Generator Tubes Pulled Out from Ulchin #1'
  11. Murakami, Y., 1987, 'Stress Intensity Factors Handbook,' pp. 204-205
  12. Cho, Y.J., 1990, 'A Study on the Interaction Effect of Adjacent Semi-Elliptical Crack,' Master's Thesis
  13. Harrison, R.P., Loosemore, K., Milne, I. and Dowling, A.R., 1980, 'Assessment of the Integrity of Structures Containing Defects,' CEGB Report, R/H/R6-Rev.2
  14. Erdogan, F., 1976, 'Ductile Failure Theories for Pressurized Pipes and Containers,' Int. J. PVP, Vol. 4
  15. Kiefner, J.F., Maxey, W.A., Eiber, R.J. and Duffy, A.R., 1973, 'Failure Stress Levels of Flaws in Pressurized Cylinders,' ASTM STP 536, pp. 461-481
  16. Majumdar, S., Shack, W. J., Diercks, D.R., Mruk, K., Franklin, J. and Knoblich, L., 1998, 'Failure Behavior of Internally Pressurized Flawed and Unflawed Steam Generator Tubing at High Temperatures - Experiments and Comparison with Model Predictions,' NUREG/CR-6575
  17. Kim, Y.J., Choy, Y.S. and Lee, J.H., 1993, 'Development of Fatigue Life Prediction Program for Multiple Surface Cracks,' ASTM STP 1189, pp. 536-550
  18. Shibata, K., Yokoyama, N., Ohba, T., Kawamura, T. and Miyazono, S., 1985, 'Growth Evaluation of Fatigue Cracks from Multiple Surface Flaws (I),' J. Japanese Nuclear Society, Vol. 28, No.2, pp. 250-262
  19. Shibata, K., Yokoyama, N., Ohba, T., Kawamura, T. and Miyazono, S., 1986, 'Growth Evaluation of Fatigue Cracks from Multiple Surface Flaws (II),' J. Japanese Nuclear Society, Vol. 28, No.3, pp. 258-265
  20. Park, Y.W., Song, M.H. and Lee, J.H., 2000, 'Steam Generator Tube Integrity Program,' KINS/RR-001
  21. Lee, J.H., Park, Y.W., Song, M.H., Kim, Y.J. and Moon, S.I., 2000, 'Evaluation of Plugging Criteria on Steam Generator Tubes and Coalescence Model of Collinear Axial Through-Wall Cracks,' J. Korean Nuclear Society, Vol. 32, pp.465-476
  22. Diercks, D.R., 2000, 'Steam Generator Tube Integrity Program Monthly Report,' ANL.
  23. Park, Y.W., Song, M.H., Lee, J.H., Moon, S.I. and Kim, Y.J., 2002, 'Investigation on the Interaction Effect of Two Parallel Axial Through-Wall Cracks Existing in Steam Generator Tube,' Nuclear Engineering and Design, Vol. 214, pp. 13-23 https://doi.org/10.1016/S0029-5493(02)00010-9
  24. Miller, A.G., 1988, 'Review of Limit Loads of Structures Containing Defects,' Int. J. PVP, Vol. 32, pp. 197-327