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David Armstrong

Dr David Armstrong
Royal Academy of Engineering Research Fellow

Department of Materials
University of Oxford
16 Parks Road
Oxford OX1 3PH
UK

Tel: +44 1865 273708 (Room 110.20.08)
Tel: +44 1865 273777 (reception)
Fax: +44 1865 273789 (general fax)

Materials for Fusion and Fission Power
Oxford Micromechanics

Summary of Interests

 

 

My work is centered around developing mechanical testing techniques at the nano and micro scale. We have a state of the art high temperature nanoindentation system which allows us to perform tests  upto temperature above 1000K. These techniques are being used to study a range of important materials for both fission and fusion power including refractory alloys, high strength steels and zirconium alloys. This is carried out with a range of partners including, UKAEA, General Atomics, USA, Rolls Royce, Karlsruhe Institue of Technology, Germany and  the University of Leoben, Austria, as well as many collaborators within Oxford. Particular areas of current research include:

 

  • Materials for nuclear fission and fusion
  • Development of micromechanical testing techniques
  • Fundamentals of fracture
  • High temperature mechanical properties
  • Time dependent deformation 
  • Ceramic composite materials

 

Current Research Projects

Micromechanical testing
Dr. D.E.J. Armstrong, Dr T.B. Britton, Dr J. Gong, Professor S.G. Roberts, Dr. A.J. Wilkinson
The project develops new methods of testing mechanical properties at the micron scale, using a combination of focussed ion beam machining (to produce specimens) and atomic force microscopy / nanoindentation (to test them). The methods are applied to testing thin films, ion-irradiated layers, interfaces and properties of individual grains and grain boundaries in alloys. A newly-commissioned machine will enable tests to be perfomed in the temperature range -50 to +750C. Supported by EPSRC and CCFE Culham (Junior Research Fellowship at St Edmund Hall: D. Armstrong)

Tungsten for fusion power applications
R. Abernethy, C. Beck, Dr. D.E.J. Armstrong, Professor S.G. Roberts, Dr, M. Reith*
Tunsten and tungsten alloys (especailly W-Ta and W-Re) are candidate materials for the "divertor" (helium exhaust) of a fusion power plant. During operation, neutron irradiation gradually transmutes tungsten into W-Re then W-Re-Os alloys. This, and the radiation danage itself, will affect its mechanical properties.  In this project, we fabricate W-Re and W-Ta alloys and subject them to W-ion irradiation, mimicking the neutron irradiation. Micromechanical test methods are used to study effects of radiation on strength. The migration of Re & Ta to grain boundaries, and how the this changes grain boundary strength, will also be studied. Funded by EPSRC and CCFE. (* Karlsruhe Institute of Technology)

Understanding Mechanical Size Effects in Oxide Dispersion Strengthened Steels
C. Jones, Dr D.E.J. Armstrong, Professor S.G. Roberts
For high temperature applications in both future fusion and fission reactors oxide dispersion strengthened steels (ODS Steels) are seen promising materials. However there is a lack of data available on how the mechanical properties (both yield strength and fracture toughness) change after neutron irradiation. Due to the time and cost constraints of neutron irradiation ion implantation can be used to mimic the neutron damage, however the damaged layer is typically only a few microns deep. Micro-cantilever bending techniques can be used to study the mechanical properties of the damaged layer, but to extract useful engineering properties the influence of size effects, due to the reduced specimen size, and how these vary with key microstructural length scales, must be understood. This project is using both Eurofer97 and ODS alloys developed in Oxford to study the size effects present in micro-cantilever tests and developing methods to allow comparisons of results from micro and macro mechanical tests.

Brittle-ductile transitions in BCC metals for fusion power applications
Dr D.E.J. Armstrong, Dr. E. Tarleton, Professor S.G. Roberts, Dr. A.J. Wilkinson, Professor S.L. Dudarev*
The project investigates the brittle-to-ductile transition in tungsten and iron-chromium alloys up to 12%Cr (all these metals are the basis for proposed fusion power plant alloys). Pre-cracked miniature bend specimens of single crystals and polycrystalline materials are fracture tested in the temperature range 77 - 450K. The effect of dislocation motion around the crack tips on fracture stress is examined, and modelled using dynamic-dislocation simulations. Funded by EPSRC and CCFE. (*EURATOM/CCFE)

Processing and properties of tungsten coatings for fusion reactors
Dr D. Armstrong, Professor P.S. Grant
Tungsten is the key plasma facing material for use in any future nuclear fusion device due to its high melting point, good sputter resistance and low activity. However its refractory nature leads to inherent difficulties in its processing and many traditional production routes are not available. Vacuum plasma spraying is one of the most attractive methods of producing tungsten coatings for this application, but thermal mis-match between the tungsten and substrates such as steel or copper lead to the development of complex residual stresses which degrade the performance of the coating. Other challenges include microstructural control, and characterising the properties of tungsten in the coating arrangement. This project will use recently upgraded vacuum plasma spraying equipment to produce both pure and alloyed tungsten coatings on novel substrates. In-situ process data will be recorded including temperatures, deflections, etc. Coatings will be characterised using state of the art microscopy and micro-mechanical testing facilities, as well as thermal loading, and finite element analysis used to understand the evolution of the stress state. By consideration of the process conditions in manufacture, the microstructure and the properties of the coatings, optimisation will be performed to produce practical tungsten coatings for further testing and application. Funded by EPSRC and CCFE

5 public active projects

Research Publications

For an upto date publication list please follow the link below 

D.E.J. Armstrong Publication List

Projects Available

Micromechanical Testing of Irradiated Nuclear Fusion Materials
Dr David Armstrong, Dr Angus Wilkinson, Dr Edmund Tarleton, Oxford, Chris Hardie CCFE

Understanding how irradiation damage from neutrons affects the mechanical properties of structural materials is a key step towards realising nuclear fusion as a sustainable power source.  However working on irradiated materials is costly, and generating mechanical data from them is difficult. Neutron damage can be simulated with ion irradiations but the damage layers are thin -  200nm to 100µm. As such traditional mechanical testing methods cannot be used and novel micro-mechanical tests must be conducted. This leads to difficulties in interpreting the results due to size effects inherent in testing small material volumes.

This project will utilise the newly opened Materials Research Facility at the Culham Centre for Fusion Energy to study the effects of ion irradiation on fusion materials and correlate this with the defect populations produced. This will then be used to develop methods to use small scale mechanical tests to aid engineering design of future fusion systems. Materials of interest include chromium, vanadium and tungsten based alloys.

 

Ion irradiations will be carried out using protons and heavy ions at a range of international irradiation facilities, at fusion reactor relevant doses and temperatures. Advanced electron microscopy at the Department of Materials, University of Oxford will be used to characterise the damage and defect types produced. Micromechanical tests will be performed at the MRF to understand how these defects affect mechanical behaviour, such as fracture toughness, work hardening, and flow localisation. Tests conducted will include nano-indentation, micro-cantilever and compression tests and micro-scale tensile tests. Finite element modelling will be used to interpret the results. This work will be in close collaboration with a defect based modelling phd based at Oxford Materials, to fully understand the mechanisms which control deformation of irradiated materials. The student will be enrolled on the Fusion CDT and the project will involve significant periods of experimental work at the MRF at CCFE as well as work in the Oxford Materials Department.

Also see homepages: David Armstrong

Modeling of Micromechanical Testing of Irradiated Nuclear Fusion Materials
Edmund Tarleton, Angus Wilkinson, David Armstrong Oxford, Chris Hardie CCFE

Understanding how irradiation damage from neutrons affects the mechanical properties of structural materials is a key step towards realising nuclear fusion as a sustainable power source.  However working on irradiated materials is costly, and generating mechanical data from them is difficult. Neutron damage can be simulated with ion irradiations but the damage layers are thin -  200nm to 100µm. As such traditional mechanical testing methods cannot be used and novel micro-mechanical tests must be conducted. This leads to difficulties in interpreting the results due to size effects inherent in testing small material volumes.

This project will involve coding, debugging and performing simulations with state of the art computer models being developed in Oxford namely a coupled 3D (DDP) discrete dislocation plasticity / finite element code and a crystal plasticity finite element code (Abaqus UMAT) to simulate nano-indentation experiments. The experiments you will simulate are being performed at the Materials Research Facility at the Culham Centre for Fusion Energy to study the effects of ion irradiation on fusion materials and correlate this with the defect populations produced. The insight gained will then be used to develop methods to use small scale mechanical tests to aid engineering design of future fusion systems. Materials of interest include chromium, vanadium and tungsten based alloys.

 

Key challenges will be how to accelerate the code using a GPU, how to implement the correct traction/displacement boundary conditions, and how to incorporate complex geometry such as multiple precipitates. You will be part of a small team developing the codes and performing simulations and will also interact closely with experimental researchers at MRF at CCFE as well as work in the Oxford Materials Department and therefore will have access to rich data sets to validate and improve the model. The ultimate goal of the project is to be able to perform virtual experiments that reproduce real experiments and in doing so fully understand the mechanisms which control deformation of irradiated materials.

Also see homepages: David Armstrong

Understanding High Temperature Small Scale Mechanical Performance of Materials for Nuclear Fusion
Dr D.E.J. Armstrong, Dr E. Tarleton, Professor A.J. Wilkinson,

Future nuclear power systems, both fission and fusion, rely on the development of materials which can withstand some of the most extreme engineering environments. These include temperatures up to 1500oC, high fluxes of high energy neutrons and effects of gaseous elements produced by transmutation and implantation from the plasmas. Due to efforts to minimise the production of nuclear waste by such reactors the elements which may be used in structural components is limited and in many cases there is a lack of understanding of the basic deformation processes occur in ether pure materials or alloys and importantly how these are affected by temperature, radiation damage and gas content. This project will build upon the expertise in the MFFP and Micromechanics groups on high temperature mechanical testing at the micro and nano-scale. Facilities include two high temperature nanoindenters (-50oC to 950oC), high temperature microhardness (RT to 1500oC) and dedicated FIB-SEM and FEG-SEM with EBSD as well as state of the art computer codes for strain gradient crystal plasticity finite element modelling and discrete dislocation plasticity modelling. Both nanoindentation, micro-compression and micro-bend experiments will be used to study plastic deformation, fracture and creep in a range of novel high temperature materials (likely Fe, SiC or Zr based) with potential for use in future fusion reactors. HR-EBSD and AFM will be used to study deformation structures produced during testing and to inform strain gradient crystal plasticity finite element and discrete dislocation models. This will allow for a fuller understanding of the underlying physics of deformation in these materials both before and after irradiation or gas implantation. Strong links will be made to activities within the Science and Technology of Fusion Energy (EPSRC Centre for Doctoral Training) and the Culham Centre for Fusion Energy.

Also see homepages: David Armstrong Edmund Tarleton Angus Wilkinson

High Entropy Alloys for Nuclear Applications
David EJ Armstrong and Angus J Wilkinson

High entropy alloys are a relatively new and unexplored class of metallic alloys in which three to five elements are mixed in near equal proportions in a single phase solid solution. The high entropy of mixing suppresses phase separation and should lead to high strength retained to high temperatures. We have demonstrated that some of these alloys can have excellent resistance to radiation damage and as such they are garnering interest as potential future fuel cladding material for fission reactors or as a structural material for fusion reactors.
This project will aim to produce a low activation alloy suitable for nuclear fusion or fission application.  Elemental powder will be combined using arc melting to produce small quantities of test alloys. Microstructures stability will be studied using thermal treatments, X-ray diffraction and SEM-EDX and EBSD. Mechanical testing will be carried out across length scales and at reactor relevant temperatures, using both nanoindentation and macro-scale mechanical testing. This will for a fundamental study of deformation processes in this relatively new class of alloys. The most promising alloys will be ion irradiated to simulate neutron damage, with TEM and micromechanical testing used to study the effect of the irradiation on mechanical behaviour. This will then lead to an understanding of the irradiation resistance of these alloys. The project may be linked to the Fusion CDT and will also link with collaborators both in the UK and USA.

Also see homepages: David Armstrong Angus Wilkinson

Strains Induced by Hydride Formation in Zirconium
Prof Angus J Wilkinson, Dr Ed Tarleton, and Dr David E J Armstrong

In service temperature cycling of nuclear fuel cladding can lead to repeated sequences of precipitation and dissolution of hydrides in zirconium based alloys. During the transformation from hydrogen in solid solution to the hydride phase there is a considerable volume expansion. This project will explore the links between nucleation sites, hysteresis between temperatures for precipitation and dissolution, the stress field and local plasticity induced by the transformation strain and the precipitation morphology. The following techniques will be used: high resolution EBSD, digital image correlation of SEM images, in situ thermal cycling, finite element based-crystal plasticity simulations. This project will be carried out in close conjunction with Rolls Royce and other partner Universities within the HexMat flagship EPSRC programme (http://www3.imperial.ac.uk/hexmat).

Also see homepages: David Armstrong Angus Wilkinson

Also see a full listing of New projects available within the Department of Materials.