Rupture Characteristics of Zircaloy-4 Cladding with Internal and External Simulation of Reactor Heating

Abstract

Rupture characteristics of internally heated Zircaloy-4 nuclear fuel cladding were investigated to develop rupture data that would provide a basis to bench-mark mathematical modeling of clad rupture. Single- and five-tube configurations of pressurized water reactor cladding were ruptured at constant pressures and heating rates with simulated reactor boundary conditions. The results showed that the α + β material property transformation region had a strong effect on the material ductility, and the clad rupture temperature varied linearly with hoop stress in the α + β region. Axial distribution of multitube ruptures was not coplanar. The rupture positions on all specimens were distributed normally about the point of maximum temperature. This paper emphasizes test techniques as well as the rupture characteristics of Zircaloy-4 clad.

In-situ tube burst testing and high-temperature deformation behavior of candidate materials for accident tolerant fuel cladding

“…The traditional tube burst testing to measure rupture stress and temperature is a wellestablished method that has helped accumulate a sizable database for fuel cladding materials [7][8][9][10][11][12][13][14]29]. In Figure 11 the rupture (burst) temperature versus hoop stress data for new alloys are compared to the database for Zircaloy-4, which is a collection of cladding rupture data for a variety of materials and pressurization conditions [29].…”

“…Traditionally, tube burst testing using high-pressure gas has been performed to measure the burst temperature as a function of internal pressure and to assess the performance of new and irradiated cladding specimens [7][8][9][10][11][12][13][14]. Typically a ~30 cm (1 foot) long tubular specimen is employed in such tube burst testing [14].…”