This thesis deals with the behavior of zirconium alloys under accident conditions in nuclear power plants. The author has created new devices and procedures for the charging zirconium samples with hydrogen and for determining the quantity of hydrogen absorbed by the zirconium. He has carried out experimental programs and determined the rate at which the zirconium cladding can absorb hydrogen at normal operational temperature, and also the speed at which the hydrogenated metal is oxidized in high temperature steam. By evaluating the results of mechanical tests he has determined the ductile-brittle transitional interval of zirconium cladding oxidized on one side, and also the threshold above which the hydrogenated zirconium can be considered brittle. Based on detailed tests he has determined the distribution of hydrogen in the zirconium components of those bundles that were simulating an earlier, real life malfunction.