G.I. Sharaievskyi
Institute for Safety Problems of Nuclear Power Plants, NAS of Ukraine, 12, Lysohirska st., Kyiv, 03028, Ukraine
DOI: doi.org/10.31717/2311-8253.25.1.3
Abstract
The nature and structure of interrelated thermophysical, physico-chemical, and materials science processes that directly affect the operational reliability parameters of fuel element claddings in non-boiling VVER-type reactors are considered. Possible causes of operational damage to the claddings of fuel elements in these reactors are analyzed, including their modern modifications made of Zircaloy‑4 alloy, produced by the American Westinghouse Electric Corporation, which have begun to be used at Ukrainian NPPs. A set of problematic issues in ensuring the nuclear safety of VVER reactors, directly associated with recorded cases of latent damage to fuel element claddings at several operating NPPs, is systematized. The characteristic manifestation of this damage is considered to be the effect of unpredictable and uncontrolled loss of fuel element cladding tightness during the operation of fuel assemblies in the active zones of VVER and PWR reactor installations. The study provides a physical rationale that the activation of corrosion processes in fuel element claddings is closely related to the processes of diffusion penetration of radiolysis products of the coolant through micropores in the protective barrier layer of zirconium oxide on the surface of fuel element claddings. It has been shown that the phase of active destruction of the base metal of the fuel element cladding corresponds to the initial formation of zirconium dioxide beneath the passive protective oxide film. The process of delamination of the protective barrier layer of zirconium oxide from the surface of the fuel element cladding, due to the accumulation of zirconium dioxide underneath it, is considered. The physico-chemical mechanisms of the development of dangerous corrosion processes in tubular zirconium fuel element claddings, which lead to the loss of their tightness, are considered. It has been shown that the reduction in the operational life of fuel element claddings in non-boiling reactors is due to the destruction and delamination of protective passive oxide films on their surfaces. The main tasks of developing promising software and hardware complexes for implementing operational automatic diagnostics of corrosion damage in zirconium alloys, in the active zones of non-boiling type power nuclear reactors based on the spectral parameters of neutron noise, are substantiated.
Keywords: non-boiling type power nuclear reactors, abnormal thermal-hydraulic regime, onset of coolant boiling, active zone, corrosion damage to zirconium alloys, operational
diagnostics of corrosion damage.
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