Coupled Thermomechanical Responses of Zirconium Alloy System Claddings under Neutron Irradiation

Abstract

Zirconium (Zr) alloy is a promising fuel cladding material used widely in nuclear reactors. Usually, it is in service for a long time under the effects of neutron radiation with high temperature and high pressure, which results in thermomechanical coupling behavior during the service process. Focusing on the UO2/Zr fuel elements, the macroscopic thermomechanical coupling responses of pure Zr, Zr-Sn, and Zr-Nb binary system alloys, as well as Zr-Sn-Nb ternary system alloy as cladding materials, were studied under neutron irradiation. As a heat source, the thermal conductivity and thermal expansion coefficient models of the UO2 core were established, and an irradiation growth model of a pure Zr and Zr alloy multisystem was built. Based on the user material subroutine (UMAT) with ABAQUS, the current theoretical model was implemented into the finite element framework, and the consequent thermomechanical coupling behavior under irradiation was calculated. The distribution of temperature, the stress field of the fuel cladding, and their evolution over time were analyzed. It was found that the stress and displacement of the cladding were sensitive to alloying elements due to irradiated growth.

Long-Term Corrosion Testing of Zy-4 in a LiOH Solution under High Pressure and Temperature Conditions

“…The alloying elements in the Zr alloy for the nuclear reactor can affect its properties It is known that tin can improve the strength of Zr, reduce its plasticity and improve its creep and corrosion [6]. The LiOH solution had at 25 • C a conductivity of 71 µS cm −1 , and its dissolved oxygen content was maintained below 2 ppm by thermal degassing at 100 • C. The pH of testing solution was about 10.5 (measured at room temperature, about 25 • C).…”

“…In this environment, the mechanical strength and corrosion behavior of the materials are of extreme importance. The candidate materials that meet such criteria, and have high thermal conductivity and low absorption cross-sections for thermal neutrons, are Zr alloys [6]. For low neutron absorption, the alloys must have no hafnium [7].…”

“…In comparison to Zircaloy-2, Zircaloy-4, which was introduced later, is a Ni free material and has better performance related chiefly to its lower hydrogen absorption under water corrosion [9]. As was presented recently in the literature, Zircaloy-4's parameters for fuel rod cladding [6] indicated a doubling of thermal conductivity for a temperature increase from 1000 to 4000 C. In the same conditions, the Young's Modulus change was only from 88 to 72 GPa. Zirlo is another alloy of Zr with less Sn and Ni, and no Cr.…”