Transactions of the Korean Society of Pressure Vessels and Piping (한국압력기기공학회 논문집)

Korean Society of Pressure Vessels and Piping (한국압력기기공학회)
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Ductility Degradation Assessment of Baffle Former Assembly Considering the Stress Triaxiality Effect

응력 삼축성을 고려한 원자로 내부구조물 배플포머 집합체의 연성저하 평가

  • Received : 2016.11.12
  • Accepted : 2016.12.22
  • Published : 2016.12.30
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Abstract

The study presents structural integrity assessment of ductility degradation of a baffle former assembly by performing finite element analysis considering real loading conditions and stress triaxiality. Variations of fracture strain curves of type 304 austenitic stainless steel with stress triaxiality are derived based on the previous study results. Temperature distributions during normal operation such as heat-up, steady state, and cool-down are calculated via finite element temperature analysis considering gamma heating and heat convection with reactor coolant. Variations of stress and strain state during long operation period are also calculated by performing sequentially coupled temperature-stress analysis. Fracture strain is derived by using the fracture curve and the stress triaxility. Finally, variations of ductility degradation damage indicator with the fracture strain and the equivalent inelastic strain are investigated. It is found that maximum value of the ductility degradation damage index continuously increases and becomes 0.4877 at 40 EFPYs. Also, the maximum value occurs at top and middle inner parts of the baffle former assembly before and after 20 EFPYs, respectively.

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References

  1. USNRC, 2010, Generic Aging Lessons Learned (GALL) Report, NUREG-1801, Rev.2.
  2. EPRI, 2010, Material Reliability Program: Development of Material Constitutive Model for Irradiated Austenitic Stainless Steels, MRP-135, Rev.1.
  3. EPRI, 2007, Material Reliability Program: PWR Internals Age-Related Material Properties, Degradation Mechanisms, Models, and Basis Data State of Knowledge, MRP-211.
  4. MSC, 2010, ANSYS User's Guide, Ver.12.1.
  5. EPRI, 2012, Material Reliability Program: Functionality Analysis for Westinghouse and Combustion Engineering Representative PWR Internals, MRP-230, Rev.2.
  6. Kim, J.S., Jhung, M.J., Park, J.S., and Oh, Y.J., 2013, "Development of User Subroutine Program Considering Effect of Neutron Irradiation on Mechanical Material Behavior of Austenitic Stainless Steels," Trans. of KSME A, Vol.37, No.9, pp.1127-1132. https://doi.org/10.3795/KSME-A.2013.37.9.1127
  7. Simulia, 2012, ABAQUS User's Manuals, Ver.6.11-1.
  8. Kim, J.S., Lee, Y.J., Jhung, M.J., and Park, J.S., 2013, "Susceptibility Assessment of Irradiation-Assisted Stress Corrosion Cracking on Lower Core Plate in Pressurized Water Reactor Internals," Proceedings of 2013 Int. Conf. on Materials and Reliability.
  9. Rice, J.R. and Tracey, D.M., 1969, "On the Ductile Enlargement of Voids in Triaxial Stress Fields," J. Mech. Phys. Solids, vol. 17, pp.201-217. https://doi.org/10.1016/0022-5096(69)90033-7
  10. Baso, Y, and Wierzbicki, T., 2004, "On Fracture Locus in the Equivalnet Strain and Stress Triaxiality Space," Int. J. of Mechanical Sciences, vol.46, pp.81-98. https://doi.org/10.1016/j.ijmecsci.2004.02.006
  11. Jeon, J.Y. Kim, Y.J., Kim, J.W., and Lee, S.Y., 2015, "Effect of Thermal Aging of CF8M on Multi-Axial Ductility and Application to Fracture Toughness Prediction," Fatigue and Fracture of Engineering Materials and Structures, Vol.38, Issue 12, pp.1466-1477. https://doi.org/10.1111/ffe.12316
  12. KHNP, Co., Ltd., Database for Reactor Internals in Domestic Operating Nuclear Power Plants.
  13. ASME B&PV Committee, ASME B&PV Code, Sec.II, Part D, 2007.
  14. Westinghouse, Design Report of Unit A Reactor Pressure Vessel Internals Components for Continued Operation, WCAP-16788-P, Rev.0, 2007.
  15. KHNP, Co., Ltd., Periodic Safety Review Report of Unit A.