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Zirconium alloy (Zircaloy-4) is widely used for fuel cladding in the Canadian Deuterium Uranium (CANDU) heavy water reactors. However, degradation due to waterside corrosion can limit the in-reactor design life of the nuclear fuel. For this reason, efforts must be made to understand the mechanisms of corrosion and to mitigate its effects. The purpose of the experimental research related to this work consists of the assessment of corrosion of the CANDU fuel cladding material. The paper presents the results obtained by in-situ monitoring of the Pressurized Heavy Water Reactor (PHWR) water chemistry effect on fuel cladding corrosion, Zircaloy-4. Optical metallographic and Scanning Electron Microscopies (SEM), as well as X-ray diffraction (XRD) analysis, were used to evaluate the corrosion behavior of the fuel cladding material, Zircaloy-4 coupons, exposed in a Primary Heat Transport System (PHTS) autoclave system. The obtained results are useful for characterizing the corrosion behavior of Zircaloy-4 coupons exposed in the autoclave system at Cernavoda NPP for long periods.
Dumitra Lucan (Sat,) studied this question.