%0 Journal Article
%T Laser Ultrasonic System for Surface Crack Visualization in Dissimilar Welds of Control Rod Drive Mechanism Assembly of Nuclear Power Plant
%A Yun-Shil Choi
%A Hyomi Jeong
%A Jung-Ryul Lee
%J Shock and Vibration
%D 2014
%R 10.1155/2014/296426
%X In this paper, we propose a J-groove dissimilar weld crack visualization system based on ultrasonic propagation imaging (UPI) technology. A full-scale control rod drive mechanism (CRDM) assembly specimen was fabricated to verify the proposed system. An ultrasonic sensor was contacted at one point of the inner surface of the reactor vessel head part of the CRDM assembly. Q-switched laser beams were scanned to generate ultrasonic waves around the weld bead. The localization and sizing of the crack were possible by ultrasonic wave propagation imaging. Furthermore, ultrasonic spectral imaging unveiled frequency components of damage-induced waves, while wavelet-transformed ultrasonic propagation imaging enhanced damage visibility by generating a wave propagation video focused on the frequency component of the damage-induced waves. Dual-directional anomalous wave propagation imaging with adjacent wave subtraction was also developed to enhance the crack visibility regardless of crack orientation and wave propagation direction. In conclusion, the full-scale specimen test demonstrated that the multiple damage visualization tools are very effective in the visualization of J-groove dissimilar weld cracks. 1. Introduction Control rod drive mechanism (CRDM) assembly includes a reactor vessel head (RVH) and many penetration nozzles made of carbon steel and alloy 690, respectively, as shown in Figure 1(a). The two dissimilar metal parts, namely, the RVH and penetration nozzle, are coupled with welding as shown in the unit structure of the CRDM assembly in Figure 1(b). The inner surface of the RVH, which is in direct contact with the primary coolant, is covered with cladding to prevent any reaction between the carbon steel ingredient in the RVH and the boric ingredient in the coolant. During a nuclear power plant (NPP) operation period, thermal and pressure loadings are concentrated on the penetration nozzles and dissimilar metal welding, which are comparatively fragile spots. As the operation period of nuclear plants has increased, there has been an increase in the growth of primary water stress corrosion (PWSCC) on the welds of dissimilar metals or penetration nozzles by cyclic stress. As shown in Figure 2(a), these PWSCCs ultimately grow into surface cracks and become the path of primary water leakage. While the boric acid ingredient in the coolant not only accumulates on the outer surface of the reactor vessel as boric acid deposits, but it also creates a cavity by reacting with the carbon steel ingredient of the RVH, as shown in Figure 2(b) [1]. Practically,
%U http://www.hindawi.com/journals/sv/2014/296426/