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Effect of Flaw Characterization on the Structural Integrity Evaluation Under Pressurized Thermal Shock

가압열충격 사고시 결함 이상화 방법이 구조물 건전성 평가에 미치는 영향

  • Published : 2001.02.01

Abstract

The reactor pressure vessel is usually cladded with stainless steel to prevent corrosion and radiation embrittlement. Number of subclad cracks may be found during an in-service-inspection due to the presence of cladding. It is specified, in ASME Sec. XI, that a subclad crack is characterized as a surface crack when the thickness of the clad is less than 40% of the crack depth. This condition is provided to keep the crack integrity evaluation conservative. In order to refine the fracture assessment procedures for such subclad cracks under a pressurized thermal shock condition, three dimensional finite element analyses are applied for various subclad cracks existing under cladding. A total of 36 crack geometries are analyzed, and the results are compared with those for surface cracks. The resulting stress intensity factors for subclad cracks are 6 to 44% less than those for surface cracks. It is proven that the flaw characterization condition as specified in ASME Sec. XI can be overly conservative for some subclad cracks.

Keywords

References

  1. 정명조, 박윤원, 이정배, 1997, 'Integrity Evaluation of Kori-1 Reactor Vessel for Rancho Seco Transient,' 대한기계학회 논문집 (A), 제21권, 제7호, pp. 1089-1096
  2. 곽동옥, 최재붕, 김영진, 표창률, 박윤원, 1999, '가압열충격을 고려한 원자로용기의 건전성 평가를 위한 결정론적 파괴역학 해석,' 대한기계학회 논문집(A), 제23권, 제8호, pp. 1425-1434
  3. Mukhopadhyay, N. K., Pavan Kumar, T. V., chattopadhyay, J., Dutta, B. K., Kushwaha, H. S. and Venkat Raj, V., 1998, 'Deterministic assessment of reactor pressure vessel integrity under pressurised thermal shock,' International Journal of Pressure Vessels and Piping, Vol. 75, pp. 1055-1064 https://doi.org/10.1016/S0308-0161(98)00109-4
  4. Keim, E., Hertlein, R., Fricke, S., Schoper, A., Termin-Morin, F. and Bezdikian, G., 1999, 'Thermal Hydraulics and Fracure Mechanics Analysis of a Small Break Loss of Coolant Accident in the French CP0-Type Reactor Pressure Vessel Integrity Assessment,' Proceedings of the ASME Pressure Vessels and Piping Conference, Vol. 388, pp. 71-77
  5. ASME Boiler and Pressure Vessel Code Section XI, 1995, 'Rules for In-Service Inspection of Nuclear Power Plant Components,'
  6. Choi, J.B., Lee, T.J., Jang, K.S., Choi, S.N., 1998, 'Effects of Thickness to Radius to Stress Intensity Factors for Internal Surface Cracks,' Proceedings of the ASME Pressure Vessels and Piping Conference, Vol. 373, pp. 511-515
  7. 장장희, 정일석, 1998, '원자로 압력용기 원주방향 용접부의 가압열충격 심사기준 온도의 적정성 평가,' 제5회 원전 기기 건전성 WORKSHOP, pp. 479-488
  8. Mohan, R., Wilkowski, G.M., Bass, R. and Bloom, J.M., 1998, 'A Study of Effects of Pipe Geometry on FAD Curves for Austenitic Stainless Steel and Ferritic Steel Piping Materials,' ASME Journal of Pressure Vessel Technology, Vol. 120, pp. 86-92
  9. Keeney-Walker, J., Bass, B.R. and Pennell, W.E., 1991, 'Evaluation of the Effects of Irradiated Cladding on the Behavior of Shallow Flaws Subjected to Pressurized Thermal Shock Loading,' Transactions of the 11th International Conference on Structural Mechanics in Reactor Technology, Vol. G, pp. 195-200
  10. 허용학, 이주진, 이해무, 1990, '압력용기소재에서의 표면균열의 형상변화,' 대한기계학회논문집 (A), 제14권 제3호, pp. 617-623
  11. 주석재, 1996, '안정피로성장 중인 표면균열 형상변화의 해석,' 대한기계학회논문집(A), 제20권 제9호, pp. 2843-2853
  12. 김진수, 최재붕, 김영진, 최성남, 장기상, 1998, '원자로의 클래드효과가 응력확대계수에 미치는 영향,' 대한기계학회논문집(A), 제22권, 제10호, pp. 1938-1946
  13. Schuster, D.J., Doctor, S.R. and Heasler, P.G., 'Characterization of Flaws in U.S. Reactor Pressure Vessels,' NUREG.CR-6471, Vol. 1, Pacific Northwest National Laboratory, 1998
  14. GRS, 1997, Reactor Pressure Vessel Pressurized Thermal Shock International Comparative Assessment Study
  15. Moran, B. and Shih, C.F., 1987, 'A General Treatment of Crack Tip Contour Integrals,' International Journal of Fracture, Vol. 35, pp. 295-310