Korea’s Recent Experience with ODSCC in Mill Annealed Alloy 600 Steam Generator Tubing

Speaker:
Joo, Kyungmun; Korea Hydro and Nuclear Power Central Research Institute; Korea

Authors:
Joo, K.; Korea Hydro and Nuclear Power Central Research Institute; Korea
Cheon, K.; Korea Hydro and Nuclear Power Central Research Institute; Korea

ID: ECNDT-0027-2018
Download: PDF
Session: Nuclear Industry 2
Room: H2
Date: 2018-06-13
Time: 13:50 - 14:10

Twenty-five nuclear power plants are in operation and five power plants are under construction in Korea. There are four power plants with steam generator (SG) tubes of Alloy 600 high temperature mill annealed (HTMA) material. Hanul Unit 3 and Unit 4 replaced the steam generator due to occurrence of the outer diameter stress corrosion cracking (ODSCC) at the steam generator support plate area. In general, Alloy 600HTMA tubes are known to be vulnerable to stress corrosion cracking. CE-type SG with Alloy 600MA tubes experience cracks in the tubesheets, top of tubesheet, tube support plate, upper freespan, and structures at the hotleg. In particular, as the number of years of operation of the nuclear power plant increases, the quantity of ODSCC generated at the region of the tube support plate increases rapidly. As a result, SGs of some power plants have been replaced early, and Hanbit 3, 4, 5, and 6 power plants are also preparing to replace their SGs. ODSCC is the biggest threat to steam generator life management. This paper introduces a cracking signal with related eddy current data and observes each crack’s characteristics.