STP1023

    Development of a Mechanistic Model to Assess the External Corrosion of the Zircaloy Claddings in PWRs

    Published: Jan 1989


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    Abstract

    Out-of-pile corrosion experiments form a part of the program implemented to progress in understanding power water reactor (PWR) fuel rod waterside corrosion. Thus, out-of-pile loop tests were performed in order to correlate the thermal/hydraulic and chemical environment of the fuel rods to the corrosion rates of the Zircaloy-4 cladding material. Comparison of corrosion data yielded by in-loop, PWR, and autoclave tests then made it possible to demonstrate the effects of heat flux on corrosion kinetics through the temperature gradient in the oxide layer. PWR corrosion data also showed that the activation energy corresponding to Zircaloy oxidation by water does not vary under neutron flux compared to out-of-pile conditions. The influence of irradiation is either a threshold phenomenon dependent on fast neutron flux or a factor related to the chemical environment (for example, inclusion of lithium by recoil effect caused by thermal neutrons). Different destructive examination techniques to characterize the zirconium oxide morphology and (Li/K) concentration profile in the oxide have allowed us to suggest a relationship between operating conditions, such as two-phase flow heat transfer, lithium hydroxide/potassium hydroxide (LiOH/KOH) levels at the metal-coolant interface, and the kinetics of the growing oxide film. Then, the corrosion model makes the following calculations possible: (1) the precise thermal/hydraulic conditions in the environment of the fuel rods in the core of a PWR; (2) the localized coolant pH and overconcentration of soluble species in the coolant under boiling regime and its influence on the waterside corrosion of the Zircaloy cladding; (3) the contribution of irradiation flux to the enhancement of the oxidation rate, and (4) the axial and azimuthal oxide thicknesses on fuel rods. Simultaneously, a study has consisted in correlating the behavior of the Zircaloy materials under different autoclave experiments, out-of-pile tests, and in-reactor conditions. The future objective is to include the metallurgical properties in the corrosion model.

    Keywords:

    Zircaloy, corrosion, thermal/hydraulic, irradiation, boiling, lithium, pH coolant, models, autoclave, loops, power reactor


    Author Information:

    Billot, P
    Fellow Engineer, head of Atomic Energy Commission, and fellow engineer, Commissariat A L'Energie Atomique, CEN Cardarache, St. Paul-Lez-Durance,

    Beslu, P
    Fellow Engineer, head of Atomic Energy Commission, and fellow engineer, Commissariat A L'Energie Atomique, CEN Cardarache, St. Paul-Lez-Durance,

    Giordano, A
    Fellow Engineer, head of Atomic Energy Commission, and fellow engineer, Commissariat A L'Energie Atomique, CEN Cardarache, St. Paul-Lez-Durance,

    Thomazet, J
    Engineering specialist, Fragema, Framatome, Lyon,


    Paper ID: STP18864S

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP18864S


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