Phd with Westinghouse in the effect of vanadium and copper on defect structure generation during neutron irradiation of zirconium-based materials

With industrial partner – Westinghouse, Sweden

In modern nuclear reactors, zirconium alloys are used for encapsulating nuclear fuel. Zirconium alloys are used because they are very transparent to neutrons, have excellent corrosion properties in a water-cooled reactor and have reasonable mechanical strength. Zirconium alloys are constantly developed further in order to utilise nuclear fuel assemblies in a reactor for longer than what is achieved today. In fact, it is the performance of the encapsulating zirconium alloys that determines how much energy can be extracted from nuclear fuel and not the enrichment level. An improvement of the so-called ‘burn-up’ will result in fewer reactor shut-downs and more power generation per unit of nuclear waste.

Nuclear reactor cores are one of the most demanding environment for structural materials. When inserted into a reactor, zirconium alloys undergo corrosion, hydrogen pick up and significant levels of irradiation damage as they encapsulate nuclear fuel. Irradiating a metal like zirconium results in very dramatic microstructural changes and therefore alterations of the performance. Hence, very detailed understanding of the microstructural evolution during irradiation is highly desirable. The irradiation-induced damage evolution also has very important consequences for the material properties such as irradiation hardening and irradiation-induced dimensional changes. The latter is of particular interest for this project as the dimensional changes are greatly affected by alloy chemistries. To date, a clear understanding of the relationship between alloy chemistry, development of defect structure during irradiation and irradiation-induced dimensional instabilities is still missing, which is a particular issue when trying to develop new Zr-alloys.

The project will focus on two development alloys, one with V and one with Cu additions, which were irradiated in the BOR-60 research reactor. X-ray diffraction-based line profile analysis will be utilised to obtain dislocation line densities for both alloys irradiated to five different fluence levels. In addition, detailed STEM analysis will explore loop arrangements and the role of Cu and V using high resolution EDX mapping. The findings will be compared with ongoing analyses of more conventional Zr-alloys irradiated during the same irradiation campaign.
The project feeds into ‘MIDAS’, a £9M EPSRC programme grant led by Manchester, with partners at Oxford, Imperial, Culham Centre for Fusion Energy and other UK and international research and industrial stakeholders, and is an exciting opportunity to join the UK’s largest fuel cladding materials programme. .

For more information please contact Prof Michael Preuss ([Email Address Removed]).