In a nuclear reactor the corrosion process between the coolant and the metal surface produces hydrogen, as well as the radiolysis of the coolant itself. The fraction of hydrogen that is picked up by the metal and diffused through, may lead to hydrogen embrittlement that affects the life cycle of the alloy. The mechanism of hydrogen absorption in nuclear materials has yet many gaps, hence, it is vital to have a better understanding of the diffusion and permeation of hydrogen in these alloys at a nuclear reactor's operating conditions. The main purpose of this research is to measure the effective diffusion coefficient (Deff) of hydrogen in zirconium fuel cladding materials at the reactor operating conditions. The challenging conditions (e.g.: high temperatures) led to the selection of a molten NaOH-KOH Devanathan-Stachurski electrochemical permeation cell as the experimental set-up, modified to use in a temperature range of 200 to 300Â°C. Zirconium (99,2%), Zircaloy-4 and ZIRLOÂ® were the designated materials to study. The outcome showed that hydrogen diffusion in zirconium and its alloys is a slow process at temperatures below 200Â°C, and hydrides were still present at temperatures of 250Â°C, with hydrogen diffusion being associated to the square of the hydrogen concentration (DeffÃÂ·CH2) with a linear relation up to 250Â°C. Moreover, the Deff calculated for zirconium at 250Â°C was 9.6 x 10-13 cm2/s, value closer to the hydrogen diffusion coefficient in the zirconium oxide barrier. The effects of the Î±-Zr and Nb-rich phases in ZIRLOÂ® is highly noted during the anodic polarisation, where two peaks associated with the hydrogen oxidation are shown and subsequently linked with the separate paths through the phases and chemistries.