| 26 | 0 | 3 |
| Downloads | Citas | Reads |
The ligament of perforated head of reactor pressure vessel head is prone to defects due to the complexity of its structure and load. In order to ensure the structural integrity of reactor pressure vessel during its service life, safety evaluation of defective parts needs to be carried out, in which the stress intensity factor is a key parameter required for evaluation. The finite element method was adopted to calculate the stress intensity factors of semi-circular cracks in the ligament of perforated head of reactor pressure vessel under pressure loading and thermal loading, respectively. The finite element results were compared with the influence function method proposed by Chapuliot. The results show that if the crack depths are small, the Chapuliot method can be conservatively used to calculate stress intensity factors. As the crack depths are larger than 0. 2 times of head thickness, the Chapuliot method is no longer suitable. In order to obtain a more accurate engineering calculation method for the stress intensity factors of cracks in the ligament of perforated head, based on the finite element solutions and the influence function method in the RSE-M standard, the influence coefficients of the central points and the surface points of the semi-circular cracks with different sizes are fitted under pressure loading and thermal loading.
[1]田宇,雷柏茂,史力,等.反应堆压力容器防断裂设计的规范法和分析法比较[J].力学季刊,2013,34(3):429-436.
[2]Oh C, Lee S, Jhung M J. Analytical method to estimate cross-section stress profiles for reactor vessel nozzle corners under internal pressure[J]. Nuclear Engineering and Technology, 2022, 54(1):401-413.
[3]Li Y, Jin T, Wang Z, et al. Engineering critical assessment of RPV with nozzle corner cracks under pressurized thermal shocks[J]. Nuclear Engineering and Technology,2020, 52(11):2638-2651.
[4]Liu R, Huang M, Peng Y, et al. Analysis for crack growth regularities in the nozzle-cylinder intersection area of Reactor Pressure Vessel[J]. Annals of Nuclear Energy,2018, 112:779-793.
[5]张丽屏,苏东川,高世卿,等.反应堆压力容器接管嘴内隅角应力强度因子计算研究[J].原子能科学技术,2017,51(11):2042-2048.
[6]王大胜,刘攀,熊光明.考虑接管载荷的反应堆压力容器接管嘴断裂力学分析[J].核动力工程,2015,36(5):120-123.
[7]Murtaza U T, Javed H M. The effects of thermal stresses on the elliptical surface cracks in PWR reactor pressure vessel[J].Theoretical and Applied Fracture Mechanics, 2015, 75:124-136.
[8]孙英学,郑斌,臧峰刚.反应堆压力容器出口接管管嘴缺陷断裂力学分析[J].核动力工程,2009,30(4):21-23.
[9]顾正军,王国珍,轩福贞,等.反应堆压力容器J形坡口焊接接头区控制棒驱动机构管座裂纹的应力强度因子[J].机械工程学报,2011,47(20):109-115.
[10]Tan J P, Zhang R K, Li Y, et al. Safety assessment of external defects in the nozzle-head intersection of a nuclear steam generator[J]. International Journal of Pressure Vessels and Piping, 2022, 199:104732.
[11]ASME B&PV Code, section XI, Rules for in service inspection of nuclear power plant components[Z]. American Society of Mechanical Engineers, 2015.
[12]API 579-1/ASME FFS-1 Fitness-For-Service[Z]. American Society of Mechanical Engineers, 2016.
[13]RSE-M. Inservice inspection rules for the mechanical components of PWR nuclear islands[Z]. French Association for Design, 2010.
[14]Juma C, Namgung I. Assessment of Equivalent Elastic Modulus of Perforated Spherical Plates[J]. Transactions of the Korean Society of Pressure Vessels and Piping, 2019,15(1):8-17.
[15]Chapuliot S. Stress intensity factor calculation in sharp and beveled edge nozzle corners[J]. International Journal of Pressure Vessels and Piping, 2016, 141:11-18.
[16]乐京霞,周恒.基于相互积分的应力强度因子数值分析[J].武汉理工大学学报(交通科学与工程版),2013,37(6):1248-1250.
Basic Information:
DOI:10.20190/j.cnki.02580918.202602004
China Classification Code:TL351.6;TM623
Citation Information:
[1]ZENG Xin,LIU Fangwei,TAN Jianping ,et al.Calculation of Stress Intensity Factor Method in the Ligament of Perforated Head of Reactor Pressure Vessel[J].Nuclear Science and Engineering,2026,46(02):265-273.DOI:10.20190/j.cnki.02580918.202602004.
Fund Information:
国家自然科学基金资助项目(No.12572088,No.52075174)
2026-04-15
2026-04-15