Document Type

Technical Report

Publication Date



The objective of this quality-affecting task is to evaluate the susceptibility of spent nuclear fuel cladding materials (zirconium alloys) to stress corrosion cracking (SCC), delayed hydride cracking (DHC) and localized corrosion (pitting/crevice) in simulated repository environments. During the first year of this project, major efforts were focused on developing the infrastructure for performing the desired testing involving two highly corrosion-resistant alloys namely, zircaloy-2 (Zr-2) and zircaloy-4 (Zr-4) in simulated concentrated acidic water (SAW) and modified SAW (SAWM). Modification of the SAW chemistry was done by adding hydrochloric acid (HC1) to achieve lower pH. The construction of the "Materials Performance Laboratory (MPL)" having numerous research capabilities was completed in May 2002. Subsequently, testing was initiated in MPL to evaluate the SCC, HE, and localized corrosion behavior of both zirconium (Zr) alloys in SAW and SAWM environments. Substantial data have been generated since then, that may be utilized in reducing uncertainties in cladding models of TSPA-LA and in addressing DOE/NRC agreements for CLST KTI, sub-issue 3. The data generated may also be used in AMR entitled "Clad Degradation - Summary and Abstraction, ANL-WIS-MD-000007."


Nevada – Yucca Mountain; Nuclear fuel claddings – Corrosion; Radioactive wastes – Storage; Stress corrosion; Zirconium alloys – Corrosion


Materials Science and Engineering | Metallurgy | Nuclear




Document Number: TR-03-010
Revision: Rev O
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