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

T12025

CRP

1828

Approved Date

5 July 2011

Status

4 - Closed

Start Date

5 July 2011

Expected End Date

4 July 2015

Completed Date

19 August 2015

Participating Countries

Argentina
Brazil
Canada
Germany
India
Japan
Lithuania
Pakistan
Republic of Korea
Romania
Russian Federation
Sweden
Switzerland
Ukraine

Description

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.

Objectives

The Objective of IAEA Project 1.2.2.1 "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

Impact

The conductance of the CRP allowed participating Member States to implement comprehensive techniques and experimental methodology to test susceptibility of Zr alloys to the Delayed Hydride Cracking which is rather serious factor assisting to loss of integrity by the water reactor zirconium-based alloys fuel rod claddings.

Relevance

Fully relevant to the IAEA project 1.2.2.1 "Nuclear power reactor fuel research and development, design and manufacture reliable and economically viable core structurals and fuels for nuclear power reactors".

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