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      응력 삼축성을 고려한 원자로 내부구조물 배플포머 집합체의 연성저하 평가 = Ductility Degradation Assessment of Baffle Former Assembly Considering the Stress Triaxiality Effect

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      https://www.riss.kr/link?id=A103049550

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      다국어 초록 (Multilingual 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 ...

      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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      참고문헌 (Reference)

      1 김종성, "중성자 조사에 따른 오스테나이트 스테인리스 강의 기계적 재료거동 변화를 고려한 사용자 정의 보조 프로그램 개발" 대한기계학회 37 (37): 1127-1132, 2013

      2 Kim, J.S., "Susceptibility Assessment of Irradiation-Assisted Stress Corrosion Cracking on Lower Core Plate in Pressurized Water Reactor Internals" 2013

      3 KHNP, "Periodic Safety Review Report of Unit A"

      4 Rice, J.R., "On the Ductile Enlargement of Voids in Triaxial Stress Fields" 17 : 201-217, 1969

      5 Baso, Y, "On Fracture Locus in the Equivalnet Strain and Stress Triaxiality Space" 46 : 81-98, 2004

      6 ASME, "Materials,”ASME B&PV, Sec.II, Part D"

      7 EPRI, "Material Reliability Program: PWR Internals Age-Related Material Properties, Degradation Mechanisms, Models, and Basis Data State of Knowledge (MRP-211), 1015013"

      8 EPRI, "Material Reliability Program: Functionality Analysis for Westinghouse and Combustion Engineering Representative PWR Internals (MRP-230, Rev.2), 1021026"

      9 EPRI, "Material Reliability Program: Development of Material Constitutive Model for Irradiated Austenitic Stainless Steels (MRP-135-Rev.1), 1020958"

      10 USNRC, "Generic Aging Lessons Learned (GALL) Report" 2010

      1 김종성, "중성자 조사에 따른 오스테나이트 스테인리스 강의 기계적 재료거동 변화를 고려한 사용자 정의 보조 프로그램 개발" 대한기계학회 37 (37): 1127-1132, 2013

      2 Kim, J.S., "Susceptibility Assessment of Irradiation-Assisted Stress Corrosion Cracking on Lower Core Plate in Pressurized Water Reactor Internals" 2013

      3 KHNP, "Periodic Safety Review Report of Unit A"

      4 Rice, J.R., "On the Ductile Enlargement of Voids in Triaxial Stress Fields" 17 : 201-217, 1969

      5 Baso, Y, "On Fracture Locus in the Equivalnet Strain and Stress Triaxiality Space" 46 : 81-98, 2004

      6 ASME, "Materials,”ASME B&PV, Sec.II, Part D"

      7 EPRI, "Material Reliability Program: PWR Internals Age-Related Material Properties, Degradation Mechanisms, Models, and Basis Data State of Knowledge (MRP-211), 1015013"

      8 EPRI, "Material Reliability Program: Functionality Analysis for Westinghouse and Combustion Engineering Representative PWR Internals (MRP-230, Rev.2), 1021026"

      9 EPRI, "Material Reliability Program: Development of Material Constitutive Model for Irradiated Austenitic Stainless Steels (MRP-135-Rev.1), 1020958"

      10 USNRC, "Generic Aging Lessons Learned (GALL) Report" 2010

      11 Jeon, J.Y., "Effect of Thermal Aging of CF8M on Multi-Axial Ductility and Application to Fracture Toughness Prediction" 38 (38): 1466-1477, 2015

      12 Westinghouse, "Design Report of Unit A Reactor Pressure Vessel Internals Components for Continued Operation" 2007

      13 KHNP, "Database for Reactor Internals in Domestic Operating Nuclear Power Plants"

      14 MSC, "ANSYS User’s Guide, Ver.12.1"

      15 Simulia, "ABAQUS User's Manuals, Ver.6.11-1"

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