Evaluation of Conditions for Hydrogen-Induced Degradation of Zirconium Alloys during Fuel Operation and Storage

Closed for proposals

Project Type

Coordinated Research Project

Project Code




Approved Date

5 July 2011

Start Date

5 July 2011

Expected End Date

4 July 2015

Completed Date

19 August 2015


Components for nuclear reactors are made from zirconium alloys because they have a low thermal neutron capture cross-section, high resistance to corrosion in high temperature water and acceptable mechanical strength.  A key to the successful operation of nuclear fuel is that the target burnup is attained before one of the several degradation mechanisms limits the life of the zirconium alloy fuel cladding. As burnup targets increase this goal becomes more onerous. During service the properties of the components can degrade because of irradiation damage, oxidation and hydrogen ingress. The consequence of the latter may be embrittlement and cracking. The IAEA plans to assist interested Member States' organizations in evaluation of hydrogen-induced degradation of mechanical and fracture properties  in different zirconium alloys, including both traditional and experimental materials, so that the degradation may be predicted and prevented. While the processes of time dependent crack propagation have been investigated in recently finished CRPs on Delayed Hydride Cracking in Zr pressure tubes and fuel cladding, this project will address the initial stages of crack development that define conditions when fuel integrity can be lost. The results from such a programme will also be relevant to evaluations of fuel transportation and dry storage.


The Objective of IAEA Project "Nuclear power reactor fuel research and development, design and manufacturing" is "to improve Member States' research and technological capabilities enabling them to use, develop, design and manufacture reliable and economically viable core structures and fuels for nuclear power reactors".

Specific objectives

Collect and analyse input data for evaluating hydrogen-induced degradation of Zr alloys in fuel storage and transportation conditions

Collect the input data for the validation of predictive models of degradation of zirconium alloys by hydride mechanisms

Contribute to the understanding of the mechanism of hydride reorientation and short-term and time dependent cracking in zirconium alloys by comparing the results obtained from a wide array of alloys

Develop and validate, through round-robin tests carried out by the participating laboratories, reproducible experimental procedures to evaluate a wide range of zirconium alloys

Measure the threshold for hydride reorientation and hydride cracking as well as the brittle-to-ductile transition in samples of each alloy with standard geometry, containing a known amount of hydrogen and tested with a jointly agreed standard method


The CRP results demonstrated importance of DHC failure mechanism in Zry-4 fuel claddings at temperatures up to ~ 300-310°C. At higher temperatures K1H value increased rapidly allowing Zry-4 to be immune from DHC. This conclusion needs confirmation for neutron-irradiated Zr-based alloys.


The project successfully addressed the very pending and difficult issue of PWR and PHWR fuel cladding integrity under in-core operation and transportation / dry storage, especially in the part of the DHC threshold conditions determination.

CRP Publications




planned for 2015 (TECDOC's Foreword and Summary are attached-see Paragraph 10)



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