Corrosion Model for Zirconium-Niobium Alloys in Pressurized Water Reactors

Abstract

An engineering model of corrosion of zirconium-niobium alloys is described. It takes account of the alloying composition, the content of lithium and boron in the coolant, the heat flux on the surface of fuel elements and the intensity of the neutron irradiation. The parametric dependences used in the model are based on the results of tests performed in autoclaves and research reactors. The results of verification of the model on data from post-reactor studies of PWR and VVER fuel assemblies operating in nominal regimes are presented.The domestically produced zirconium-niobium alloys E-110 and E-635 are ordinarily used as structural materials in VVER fuel assemblies. The alloys M5 (Zr-1%Nb-Fe) and ZIRLO (Zr-1%Nb-Sn-Fe) are used in PWR fuel assemblies ( Table 1). The alloys E-110opt and E-635M based on zirconium sponge are under consideration as structural materials for Russian made fuel assemblies in the KVADRAT fuel assemblies.The main differences in water chemistry between PWR and VVER are the use of LiOH instead of KOH and dosing hydrogen instead ammonia in the coolant. Water chemistry based on LiOH can give rise to accelerated corrosion of zirconium alloys [2]. For this reason, research for additional validation of the corrosion behavior of the modified alloys E-110 and -635 is being conducted abroad for licensing fuel.Engineering models of the corrosion of zirconium alloys are used for validation of fuel serviceability [2-4]. These models are used to predict the thickness of the oxide layer and the hydrogen content in the cladding. In 2010-2012, an engineering model of the corrosion of E-110 and -635 alloys and their modifications was developed at the Bochvar All-Russia Research Institute for Inorganic Materials (VNIINM). The model is used in the computational fuel software making it possible to take account of the evolution of the height distribution of the temperature in the fuel-element cladding at the oxide-metal interface and the dependence of the corrosion rate on the heat flux density on the surface of fuel-element cladding. The present article describes the model and presents some results of verification.General Form of the Engineering Model. The engineering model was constructed using an approach similar to foreign models [2,3]. The stages of oxidation before and after the turning point are examined. The model is based on the Arrhenius relation for the corrosion rate in pure water as a function of temperature. Additional factors take account of irradiation, water chemistry and the concentration of the alloying additives. Each factor represents a parametric dependence,

Effect of Bonding Temperature on Microstructure and Mechanical Properties of 304L/Zircaloy‐4 Diffusion‐Bonded Joints with Ni/Ta Hybrid Interlayer

“…[1,2] Zircaloy is the common material in the in-core components in the nuclear industry, such as in fuel cladding in pressurized water reactors, grids, and pressure tubes in lightand heavy-water nuclear reactors, because of its inherent low neutron absorption across the sections, excellent corrosion resistance, and high temperature mechanical properties. [3][4][5][6] However, the application of zircaloy is limited due to its high cost. To reduce the production cost, the out-of-core components are usually made of stainless steel, (SS) which has favorable mechanical properties, good machinability, corrosion resistance, and low cost.…”

Long-Term Oxidation of Zirconium Alloy in Simulated Nuclear Reactor Primary Coolant—Experiments and Modeling

“…As a next step of model verification, its predictions are compared with data on oxide thickness distribution by fuel rod height after three and six years of service [ 31 ] in a WWER-1000 reactor ( Figure 14 b), using the appropriate temperature distribution [ 32 ]. The quality of prediction is quite good for three years of service and becomes somewhat worse after six years.…”