(Received 20 January 2016; accepted 15 June 2016)
Published Online: 26 August 2016
CODEN: MPCACD
  | Format | Pages | Price |   |
![]() |
PDF (2.42 MB) | 14 | $25 | ![]() |
Cite this document
The alloy 800 is being used for steam generator tubings for pressurized heavy-water reactors (PHWRs). This material is susceptible to degradation because of severe operating conditions, like high temperature, stress, and corrosive environment. These degradation mechanisms include primary water stress corrosion cracking (PWSCC), secondary side or outer diameter stress corrosion cracking (ODSCC), inter-granular corrosion (IGC), fretting, wear, denting, high cycle fatigue, corrosion fatigue, etc. The present study was conducted to analyze the effect of solution annealing temperatures, sensitization treatments, and surface conditions on the corrosion rate. The alloy 800 tubular samples were investigated by means of a conventional corrosion test (according to ASTM G28-02) and an electrochemical potentiokinetic reactivation (EPR) test. Susceptibility to inter-granular corrosion under various experimental parameters was examined by using both the test methods and the results are compared. The observed trends in corrosion rate obtained by using conventional method and EPR test were found to be similar. These results were used to obtain most optimum heat-treatment parameters and surface condition, which will yield best corrosion resistance.
Author Information:
Nayak, I. K.
Nuclear Fuel Complex, Hyderabad,
Rao, S. V. R.
Nuclear Fuel Complex, Hyderabad,
Kapoor, K.
Nuclear Fuel Complex, Hyderabad,
Stock #: MPC20160003
ISSN:2165-3992
DOI: 10.1520/MPC20160003
Author