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Fatigue Evaluation for the Socket Weld in Nuclear Power Plants  

Choi, Young Hwan (Korea Institute of Nuclear Safety)
Choi, Sun Yeong (Korea Atomic Energy Research Institute)
Huh, Nam Soo (School of Mechanical Engineering, Sungkyunkwan University)
Publication Information
Corrosion Science and Technology / v.3, no.5, 2004 , pp. 216-221 More about this Journal
Abstract
The operating experience showed that the fatigue is one of the major piping failure mechanisms in nuclear power plants (NPPs). The pressure and/or temperature loading transients, the vibration, and the mechanical cyclic loading during the plant operation may induce the fatigue failure in the nuclear piping. Recently, many fatigue piping failure occurred at the socket weld area have been widely reported. Many failure cases showed that the gap requirement between the pipe and fitting in the socket weld was not satisfied though the ASME Code Sec. III requires 1/16 inch gap in the socket weld. The ASME Code OM also limits the vibration level of the piping system, but some failure cases showed the limitation was not satisfied during the plant operation. In this paper, the fatigue behavior of the socket weld in the nuclear piping was estimated by using the three dimensional finite element method. The results are as follows. (1) The socket weld is susceptible to the vibration if the vibration levels exceed the requirement in the ASME Code OM. (2) The effect of the pressure or temperature transient load on the socket weld in NPPs is not significant because of the very low frequency of the transient during the plant lifetime operation. (3) 'No gap' is very risky to the socket weld integrity for the specific systems having the vibration condition to exceed the requirement in the ASME OM Code and/or the transient loading condition. (4) The reduction of the weld leg size from $1.09*t_1$ to $0.75*t_1$ can affect severely on the socket weld integrity.
Keywords
nuclear power plant; piping; socket weld; fatigue; integrity;
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  • Reference
1 American Society of Mechanical Engineer, ASME Boiler and Pressure Vessel Code Sec. XI, Rules for In-Service Inspection of Nuclear Power Plant Components (1998)
2 Operation Procedure for Primary Sampling System, Yong-Kwang Unit 3&4 (1995)
3 USNRC, 'Investigation and Evaluation of Cracking Incidents in Piping in PWR', NUREG-0691 (1980)
4 S. Y. Choi and Y. H. Choi, Piping Failure Analysis for the Korean Nuclear Power Piping including the Effect of In-Service Inspection, APCNDT 2003, Jeju (2003)
5 ASME Code OM, Requirements for Pre-operational and Initial Start-Up Vibration Testing of Nuclear Power Plants, ASME Boiler and Pressure Vessel Code Operation and Maintenance(OM) Part 3 (2001)
6 EPRI,Nuclear Reactor Piping Failures at US Commercial LWRs: 1961 - 1997, EPRI TR-110102 (1998)
7 USNRC, Rate of Initiating Events at U.S. Nuclear Power Plants: 1987 - 1995 NUREG/CR-5250 (1998)
8 USNRC, 'Erosion/Corrosion-Induced Pipe Wall Thinning in US NPPs', NUREG-1344 (1989)
9 OPDE Database(rev. 0.a), OECD Piping Failure Data Exchange(OPDE) Project, OECD/NEA (2002)
10 USNRC, 'Pipe Cracking Experience in LWR', NUREG 0679 (1980)
11 ASME, Class 1 Components, ASME Boiler and Pressure Vessel Code Sec. III NB (2001)