Properties of Reactor Fuel Rod Materials at High Temperatures: Final Summary Report--Severe Core Damage Property Tests Program.

摘要

The report summarizes work sponsored by the U.S. Nuclear Regulatory Commission Division of Accident Evaluation to investigate those physical properties that are needed to predict the behavior of fuel-rod assemblies during a loss-of-coolant accident. The results include a determination of the oxidation kinetics of Zircaloy and Zircaloy-uranium oxide mixtures in steam and steam-hydrogen gas mixtures at 1300 to 2400 C, viscosity measurements of zirconium-oxide mixtures at 1800 to 2100 C, an estimate of the heat of reaction for the dissolution of uranium oxide by molten zirconium at 2000 C, and thermal diffusivity measurements on prereacted Zircaloy-uranium oxide mixtures at 800 to 1500 C.