The degradation of alloy 600 and its weld material (alloy 82/182) has been reported in many nuclear power plants. In Korea, the crack induced by PWSCC was discovered in the drain nozzle of Yongkwang units 3 & 4 in 2006∼2008 and SG plug weld of Yongkwang unit 3 in 2007. In July 2007, during visual inspections of SG tube plugs at Yonggwang unit 4, boric acid deposits were observed around five Alloy 600 welded plugs. The root cause of the cracking in alloy 600 plugs was revealed to be due to the fact that the cracks were mainly caused by residual stress induced from the welding, expanding and tight-fitting. Younggwang unit 3 found the white small deposits on the drain nozzle on the 10th RFO in 2007. The root cause of the cracking in drain nozzle was revealed to be due to the initiation of a crack on the inside surface of drain nozzle and propagated to through wall cracks in the axial and circumferential direction. Younggwang unit 3 found the white widespread deposits on the upper head of a reactor vessel on the 12th RFO in 2010. Utility is trying to reveal the root cause of the cracking in the vent line of the reactor head according the KINS requirement. In this article, Korean regulatory experiences for PWSCC are introduced. After these PWSCC experiences, all SG tubes welded by Alloy 600 were replaced and all SG drain and instrumentation nozzles with Alloy 600 have been replaced into Alloy 690 material.

As the design life of new nuclear power plant increases, the austenitic stainless cladding integrity of reactor vessel becomes one of the new concerns. Since 1970’s, there have been some specific recommendations on delta ferrite content of austenitic cladding of reactor vessels and welds. It has been known that the delta ferrite is beneficial for reducing micro-fissure in welds, though the high delta ferrite content increases the probability of embrittlment of welds. In this study, the mechanical and microstructural properties of austenitic weld metals with the limit values of the recommended range (5 ∼ 18 FN) of the delta ferrite control on low alloy steels were characterized by using bending test and scanning electron microscopy. The base metal was ASME Code Sec. II specification SA 508 Gr. 3 Cl. 1 plate and weld materials were EQ308L and EQ309L strips. Four kinds of cladding were deposited with submerged arc welding process on SA508 cl.3 plates. The bending tests were performed through ASME code Sec. IX and the microstructure of fractured surfaces was analyzed by scanning electron microscopy (SEM). In bending tests, there were no fractures except the highest delta ferrite content specimens (28FN). From the SEM observation of fractured surfaces, cracks initiated from the interface between austenite and ferrites phases in the cladding layer and propagated through the continuous interfaces between two phases. For specimens without continuous interfaces of two phases, though the cracks were observed in the interface of phases, the propagation of cracks was not observed. From the test results, continuous interfaces between austenite matrix and ferrite phase provide the path for crack propagation. And the delta ferrite content affects the integrity of cladding of reactor vessel.

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