This test method is intended to include a concise description of an orderly procedure for conducting high-temperature oxidation of samples of zirconium-based nuclear fuel cladding in a steam atmosphere, followed by quenching in water and determination of post-quench ductile-to-brittle transition using ring-compression tests (RCT). These testing conditions are related to postulated loss-of-coolant accidents (LOCA) in nuclear reactors. This test method may include procedures or recommendations for hydrogen charging of zirconium-based cladding, given that hydrogen content is a significant factor on the ductility of the material.
The U.S. Nuclear Regulatory Commission (NRC) is preparing a Regulatory Guide (Draft Regulatory Guide DG-1262) describing an experimental technique for measuring ductile-to-brittle transition for zirconium-based cladding. The purpose is to prevent cladding embrittlement during a loss-of-coolant accident (LOCA), thus ensuring that the general core geometry will be maintained and be coolable. The experimental technique currently proposed by the NRC is based on the experience of a single laboratory and a limited range of test parameters, and may impose acceptance criteria disregarding concerns of the nuclear power industry on uncertainty, repeatability of results, and experimental difficulties. For this reason, the nuclear power industry is aiming to establish an ASTM (i.e. a consensus) standard test method to conduct these tests, for which a round robin program has been started. The standard test method will be used by nuclear fuel vendors, zirconium alloy vendors and contracted laboratories to test different zirconium-based cladding materials and assess compliance with the NRC regulations.
Keywordspost-quench ductility; zirconium cladding; LOCA
The title and scope are in draft form and are under development within this ASTM Committee.Back to Top
Draft Under Development