Potential Corrosion and Degradation Mechanisms of Zircaloy Cladding on Spent Nuclear Fuel in a Tuff Repository

doi 10.2172/59331

Full Text

Abstract

Available in

Date

September 1, 1984

Authors

Publisher

Office of Scientific and Technical Information (OSTI)

Related search

Decontamination of Zircaloy Cladding Hulls From Spent Nuclear Fuel

Journal of Nuclear Materials

High Energy Physics

Materials Science

English

Carbon-14 in Waste Packages for Spent Fuel in a Tuff Repository

Materials Research Society Symposium - Proceedings

Mechanics of Materials

Materials Science

Condensed Matter Physics

Mechanical Engineering

English

Ductile-Brittle Behavior of Zircaloy Fuel Cladding.

English

Corrosion of Spent Nuclear Fuel: The Long-Term Assessment

English

Effect of Chemical Species in Spent Nuclear Fuel Reprocessing Solution on Corrosion of Austenitic Stainless Steel

Transactions of the Atomic Energy Society of Japan

English

Preliminary Drift Design Analyses for Nuclear Waste Repository in Tuff

International Journal of Rock Mechanics and Mining Sciences & Geomechanics Abstracts

English

Ecohydrological Responses to Diversion of Groundwater: Case Study of a Deep-Rock Repository for Spent Nuclear Fuel in Sweden

Environmental Chemistry

English

Advanced Three-Dimensional Thermal Modeling of a Baseline Spent Fuel Repository

English

Stress Corrosion Cracking Tests on High-Level-Waste Container Materials in Simulated Tuff Repository Environments

English