The concept of coating the currently used nuclear fuel cladding (zirconium-based alloy) with an oxidation preventive layer could be a solution to suppress undesirable fast reaction kinetics with high-temperature steam. Critical Heat Flux (CHF) is a thermal-hydraulic performance parameter to investigate for these new cladding concepts including Cr-coated Zirc-based cladding. The exposure of coated Zirc-based cladding for an extensive period inside the reactor to severe conditions of high pressure and temperature, water chemistry, and irradiation environment could result in structure and property changes. These changes would have an impact on the heat transfer performance parameters including CHF. Surface wettability analysis, including contact angle (?) and surface roughness (Ra), are simple surface measurements used in the literature to predict changes in the pool boiling heat transfer coefficient and CHF. This work presents the effects of ion irradiation on surface wettability measurements and predicted pool boiling CHF for bare Zirc-4, and multiple Cr-coated Zirc-based samples of different coating thicknesses (5 - 30 ?m). The Scanning Electron Microscope (SEM) images show no significant surface damages due to ion irradiation. The ion irradiation shows no effects on the measured surface roughness for all samples. Results show that decreasing the Cr-coating thickness results in a higher surface wettability pre- and post- ion irradiation. The irradiated Cr-coated surfaces have higher measured contact angle compared to the unirradiated surfaces. The unpolished coated Zirc samples have larger measured surface roughness compared to the polished surfaces pre- and post- ion irradiation. The predicted pool boiling CHF using the contact angle measurements implemented in Kandlikar model for unirradiated samples is higher than those for irradiated samples.