Open AccessAnalysis for Materials Test Reactor (MTR Fuel Assemblies in Dry Storage)
Author(s) -
Robert Miller
Publication year - 1999
Publication title -
osti oai (u.s. department of energy office of scientific and technical information)
Language(s) - English
Resource type - Reports
DOI - 10.2172/14488
Subject(s) - corrosion , cladding (metalworking) , materials science , spent nuclear fuel , metallurgy , spent fuel pool , creep , relative humidity , fission products , fuel tank , alloy , zirconium alloy , aluminium , nuclear engineering , composite material , radiochemistry , chemistry , physics , engineering , thermodynamics
This report documents a creep analysis to estimate the maximum acceptable temperature for spent aluminum clad nuclear fuels in dry storage
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