During vitrification of nuclear high level waste, the molten glass containing the waste oxides is poured into 304L SS canisters at 1323 K (1050 degrees C), which subsequently, cools through the sensitization temperature, wherein, nucleation of chromium carbides occur. Due to the radioactive decay heat, the canister is exposed to 373 K-573 K (100 degrees C- 300 degrees C) over long containment periods, leading to the growth of the carbides, resulting in low temperature sensitization (LTS). The sensitized microstructure is susceptible to intergranular corrosion (IGC) when it comes in contact with ground water. In the present work, AISI Type 304L SS containing 0.02% and 0.03% carbon were subjected to LTS, simulating 10 and 100 years of vitrified nuclear waste containment respectively, and subsequently, the corrosion resistance was evaluated in simulated ground water. The micro structural evaluation was carried out by optical, scanning electron microscope (SEM), transmission electron microscope (TEM) and electron backscatter diffraction (EBSD). Double loop electrochemical potentiokinetic reactivation (DL-EPR) experiments were conducted to quantify the degree of sensitization (DOS). The study revealed that 304L SS containing 0.02% C is resistant to LTS and IGC, while that containing 0.03% C is highly susceptible to LTS and IGC in ground water composition.