Published: Jan 1977
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The U.S. Atomic Energy Commission (AEC) Acceptance Criteria for Emergency Core Cooling Systems stipulated conservative positions with respect to the transient temperature limits of Zircaloy because of the scatter or nonstatistical nature of existing data. In the United States and in a number of other countries, programs were initiated to obtain the desired information. The results of U.S. Nuclear Regulatory Commission (NRC) sponsored new research on Zircaloy oxidation kinetics, expansion and rupture, transient radiation damage annealing, and oxygen diffusion in β-Zircaloy are presented which serve to quantify the conservative positions taken. Additional short-term data needs are presented. Expected trends of future research are indicated to be in the direction of improved understanding of cladding failure from pellet-cladding interaction during power ramps or anticipated transients without scram (ATWS) and in understanding (experimentally and analytically) the multiaxial yielding of anisotropic materials such as Zircaloy.
zirconium, zirconium alloys, nuclear fuel cladding, safety, steam oxidation, deformation, oxygen diffusion, light water reactors
Chief, U.S. Nuclear Regulatory Commission, Washington, D.C.,
Paper ID: STP35561S