Materials for fusion & fission power

Lead Research Organisation: University of Oxford
Department Name: Materials


It has been predicted that by 2012 the UK's electricity generating capacity will no longer be enough to meet demand. Reliable new sources of multi-gigawatt electrical power will be vital for social stability and economic strength. Nuclear fusion and advanced fission power plants have been proposed, with possible years for operation in the range 2025 (advanced fission) to 2050 (fusion). These have the potential for large-scale, clean, CO2-free power generation for generations. However, they will not be viable unless some very difficult materials science problems are solved. The structural materials from which the power plants' core components will be built must have high strength and toughness at high temperatures, and retain good properties for decades despite being subjected to radiation damage from high-energy neutrons. The neutrons knock atoms from their positions, scrambling the materials' carefully-designed microstructures, and produce many small crystal defects which make the materials more brittle. The neutrons, unlike those in current nuclear power plants, have enough energy to cause transmutation reactions: this causes two problems. First, many elements ordinarily used in strong alloys cannot be used, because their transmutation products are highly radioactive for thousands of years, so we must design new strong alloys using a very restricted range of elements. Second, helium is produced in most reactions, and adds to the embrittling effects of the radiation damage.There are no fast-neutron facilities, and even slow-neutron test reactors are very expensive to use and take years for a single run . To develop the critical new materials quickly, we need to act now. We can use computer modelling of how the radiation-induced defects are formed, how they behave and how they interact to change material properties. Experimentally, ion irradiation can be used to produce the same damage types as from fast neutrons, in a few hours and without producing hard-to-handle radioactive specimens; but the amount of material affected is tiny - a layer 1/1000 mm thick. We have developed new techniques to test specimens made in these thin layers, and can use advanced microscopy to look at the radiation damage. This project will develop modelling and experiment further, and use them together so that experiments provide information to models and test their predictions. Researchers at Oxford, Liverpool and Salford Universities, UKAEA Fusion and the CEA will work together in a large project to form specialist small research teams developing innovative modelling and experimental methods, working on a problems critical to the applications of new alloys of steel and tungsten: how radiation damage can concentrate some elements at grain boundaries, making them brittle; how radiation effects on nanometre-sized oxide particles included in the alloys for high-temperature strength and to soak up helium and hydrogen.The project will make major advances in innovative experimental and modelling techniques operating at the microstructural scale where materials properties are determined, and it will verify the models' predictions against experimental data. Its success will significantly speed development of the new materials that are essential for the commercial realisation of fusion and new-generation fission power. It will help the UK to lead scientific developments in new materials and to train future experts for future fission and fusion programmes. The developments are also relevant to other important structural integrity issues (e.g. embrittlement, ductile-brittle transitions, stress corrosion cracking, and alloy strength). The project's leaders currently head world-leading research efforts in the areas which will form this integrated project. They are well-linked into the international fusion and UK fission communities, representatives of which will advise on the programme's direction and will speedily implement its results.

Planned Impact

Looking ahead 20-30 years, this project addresses issues that will be show-stoppers for the exploitation of power sources that are two of the very limited options available to the post-oil world; everyone and their descendants are thus potential beneficiaries. This project alone cannot, of course, solve these national and global problems, but it will make a substantial contribution. It will also train new experts, maintain the UK's position in worldwide fusion materials research and will link the present rather separate fusion and fission materials research communities. The lead researchers and the steering committee members are all embedded in a wide range of other research projects, collaborations and committees. At the policy-influencing and forming level, Smith (Royal Society, DBERR, IOM3), Dudarev (EFDA), English (NNL) are highly active, and our collaborations with UKAEA, CEA and RR closely involve policy-level staff. In Oxford, the project will engage with the James Martin 21st Century Institute and the Smith School of Enterprise & the Environment, which have growing external visibility and influence. UK industry will gain from the project. To the companies and institutions collaborating in the project, the longer-term benefits are simple: unless the materials issues are solved, future reactors in which these bodies have a large stake either simply will not work, or will not work for long enough to be commercially viable. For current and next-generation fission plant, the project will tackle lifetime-determining radiation-induced materials degradation. If near-term exploitable results are identified of commercial value to one or more partners or external bodies, exploitation arrangements will be made, most probably, via Oxford's technology transfer subsidiary ISIS innovation. The further development in this project of the Micro-materials nanoindenter system will help raise this UK company's profile world-wide, and open up new markets for the company. Our collaborators, UKAEA, Culham, the CEA, Rolls Royce and Corus are representative of a wider range of companies and research institutes in the fission, fusion and materials production areas, all of whom will benefit from improved knowledge of radiation effects on advanced nuclear materials and in the development of the novel materials to be studied in this programme. The lead researchers' and our collaborators' wide connections, and the workshop series we will organise, will allow us to identify possible future partners and users and to identify new potential exploitation and application routes. The project website will keep the general public and potential users of our work up to date with developments and will give clear routes for communicating with us, whether for speakers for school or science club meetings or for discussions with interested researchers or companies. The training aspects of the programme and its role in developing research careers are important. One of its significant outputs for the UK will be scientists trained in nuclear materials. Our collaborators will also benefit by participating in the training of potential future employees. This flagship project will attract new blood into nuclear materials science, and transfer knowledge, skills and experience from a rapidly-aging researcher-base into the next generation.


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Description This brief summary of the major achievements of the MFFP project since starting in 2009 is largely confined to the original target areas of the MFFP project, in ODS steels and tungsten alloys for extreme nuclear environments. However the "greater MFFP group" now includes a substantial number of researchers on other nuclear materials topics, such as hydrogen pick-up in and oxidation of zirconium alloys, stress corrosion cracking in stainless steels, fracture of nuclear graphite, bubble-lattice formation, and irradiation effects in beryllium.
1. Mechanical Alloying: We have established, from scratch, a facility for production of high-quality ODS alloys in small experimental batches, to enable optimisation of production routes and eventual alloy properties, and to provide a source of alloys of known, tailored compositions and microstructures for research within and outside the project.
2. Liquid phase routes: We have made first steps towards validating possible bulk production method, using spray forming and rapid solidification routes.
3. Additive Manufacturing: In collaboration with AMES labs, we have been processing, with some success, gas atomized reaction synthesized (GARS) powders by Selective Laser Melting (SLM) to evaluate whether solid structures can be built by laser processing of GARS powders, which also contain a dispersion of oxide particles.
4. Recovery and Recrystallisation Behaviour of ODS Powders: The recovery and recrystallisation behaviour of mechanically alloyed powders, used to make commercial FeCrAlY ODS alloys PM2000, MA956 and ODM751, has been investigated, in order to follow the nucleation and growth of oxide dispersoids at each stage during mechanical alloying of the powders and their subsequent consolidation and heat treatment.
5. Joining: In conjunction with TWI, we have now successfully produced dissimilar metal welds between PM2000 ODS steel plates and plates of other high temperature alloys (such as Kanthal APMT). The dispersoids coarsened slightly during welding, and exhibited different size distributions at different points within the weld zone.
1. ODS characterisation: High spatial resolution characterisation by APT and TEM has been performed throughout the project, of the new batches of ODS material produced by mechanical alloying at Oxford, to allow the processing conditions to be optimised. Several new types of nanoscale oxide particles have been found and the incorporation of minor elements within the oxide particles has been studied.
2. Helium management in ODS alloys: TEM studies of helium implantation in to ODS steels has shown that alloys processed at lower temperatures than the "standard" 1150°C, so as to have a greater density of ODS particles, have increased resistance to bubble formation in the bulk and at grain boundaries.
3. Radiation Damage characterisation: extensive post-irradiation TEM and TEM in-situ irradiation of tungsten, tungsten alloys and model iron-chromium binary alloys has been used to build a comprehensive matrix of the densities and types of dislocation loop damage produced over temperatures from 300°C to 800°C. Particularly notable results include:
a. Displacement damage in tungsten and tungsten alloys: Extensive bulk and in-situ irradiations have led to: a clustering scaling law for individual cascades (at low temperatures: 30 K) and evidence of sub-branching of cascades at PKA energies > 150 keV; the identification of ½<111> interstitial loop dominated microstructures at intermediate and high doses and their evolution towards ordered structures (strings, complex networks); the clarification of the influences of alloying elements (Re, Ta, V) and impurity atoms (C), irradiation temperature, dose, dose rate and geometry (grain orientation).
b. High-temperature stability of damage in tungsten: post-irradiation annealing of bulk specimens and in-situ thin-foils (isochronal, isothermal and dynamic annealing) has established the evolution path of radiation damage from R.T. to 1400°C: isolated loop hopping -> large scale spatial ordering of loops, coalescence of loops -> loops break up into lines, line-loop interaction -> surface absorption. An activation energy of ~ 1.34 eV was derived for the annealing of dislocation length in the 700-1100°C range. Interstitial clusters are most stable in the form of ½<111> pure edge loops, while vacancy clusters adopt void configurations at high temperatures (up to 1400°C).
c. Helium irradiation in tungsten and alloys: We have found that He plays an important role in the nucleation of interstitial loops; He-V-SIA complexes possibly act as the nuclei. In tungsten irradiated at 500°C, loops are largest (and lowest in density) with He+ irradiation, followed by W+ irradiation and smallest (and highest density) for dual-beam irradiation. These effects are increased by the presence of alloying atoms (Re, Ta, V). At 800°C, the critical dose for damage saturation increases.
d. FeCr alloys: TEM in combination with APT has shown that Cr segregates to dislocation loop damage, especially at lower dose rates and at lower temperatures, causing enhanced hardening.
e. ODS steels: in-situ TEM ion irradiation has demonstrated the extraordinary stability of these materials to microstructural change and damage accumulation up to damage levels of at least 10dpa.
f. FIM: Field Ion Microscopy combined with layer-by-layer tip evaporation has been applied to the study of radiation damage in tungsten, and we have developed new image analysis methods to treat the data thus derived. We have demonstrated that individual point defects from radiation damage can be imaged.
g. Phase decomposition under irradiation in W alloys: We have demonstrated that even thermodynamically stable dilute W-Re, W-Ta and W-Re-Os alloys can form extensive clusters during irradiation, leading to considerable hardening.
h. Thermal annealing of damage in irradiated W: In-situ and ex situ TEM, static anneals and dynamic temperature ramp-up at fusion reactor relevant conditions show a dramatic acceleration of loop loss above 900°C, and an activation energy for annealing of dislocation length of 1.34±0.2 eV was determined in the 700-1100°C range.
1. Micromechanical testing: Test techniques based principally on microcantilever bending have been developed and used through the MFFP project.
a. We have developed data analysis methods based on iterative matching of FEA models of beams to extract elastic, yield, work hardening / softening and fracture strength characteristics reliably from test data.
b. We have demonstrated the use of microcantilever testing on neutron irradiated materials, for the first time using these methods to comparing flow at this scale with that in equivalent ion-irradiated material.
c. Interpretation of micromechanical yield and flow data in terms of bulk properties is made difficult by strong inherent size effects in almost all materials (ODS steels may be an exception). Ideally specimens of at least 4-5?m in section are needed. This places limitations on ion energies that can be most effectively used for self-ion implantation, especially in tungsten. However, in neutron irradiated materials (or proton-irradiated materials, not yet attempted in the MFFP project), cantilevers of this size can straightforwardly be made, given enough FIB time.
d. Fracture toughness measurements can reliably be made by microcantilever bending tests if materials have KIc less than about 10MPa?m. These can be applied to measure grain boundary and interfacial strength properties, demonstrated in tungsten alloys and in oxidised Ni-600.
e. While microcantilever (or micropillar) methods have unique utility for investigation of detailed flow behaviour, they should be used in conjunction with simple nanoindentation tests to give "screening" modulus and hardness data, which can be obtained in bulk without the need for scarce FIB resources.
2. High- and low- temperature micromechanics: Commissioning the system occupied much longer than planned. We have now demonstrated the stability of the system up to 800°C, and down to -50°C. We have used it to study high temperature behaviour of He-irradiated tungsten, and have found that the very high hardening effects produced (and by extrapolation, embrittlement, though not yet tested extensively) are retained up to 800°C. The hardening centres, probably He-vacancy clusters, are too small to resolve by TEM.
The system has also been used to study carbon-dislocation interaction in iron, in the first-ever study of dynamic strain aging by indentation methods, and high temperature mechanical behaviour of molybdenum.
1. Dislocation dynamics modelling of micromechanics: A 2D discrete dislocation plasticity (DDP) model for micro cantilevers has been developed and predicts the response of zirconium and titanium beams orientated in plane strain and showed the importance of the source density in size effects. The code has now been extended to include cohesive elements to simulate plasticity and fracture concurrently. Simulations are currently underway to simulate notched beams.
2. 2D dislocation modelling: we have developed a model for an infinite isotropic elastic solid (periodic boundary conditions), to allow relaxed dislocation structures to be simulated. Predictions have been verified experimentally by EBSD and ECCI.
3. 3D dislocation modelling: we have investigated the effects of image forces important in thin films such as TEM foils. We have developed a novel simulation code harnessing the power of GPU processing, achieving orders of magnitude efficiency improvements for a fraction of the cost of a cluster. We have written a 3D FEM code and coupled it with DDLab to allow cantilevers to be simulated in 3D. Currently the code is being optimised.
4. Stochastic dislocation dynamics: we have developed an algorithm for stochastic dislocation dynamics, taking into account the fluctuating thermal forces that govern microstructural evolution at high temperatures. This has been implemented in a mesoscale 3D dislocation dynamics code, and the diffusivity of prismatic loops has been benchmarked to molecular dynamics simulations. The effect of irradiation on climb forces has been explicitly investigated at the mean-field level and compared to in situ TEM observations of prismatic loops growth in FeCr and W foils.
5. Effect of elastic anisotropy on microstructure: ??Fe and ferritic steels become highly elastically anisotropic as the ?-? transition temperature is approached. Our modelling has established the causal link between the anisotropy, the emergence of atypical sharp-cornered dislocation structures and the well-documented loss of strength at high temperatures. The jagged microstructure at the centre of the effects has been quantitatively confirmed in TEM.
6. Grain boundary helium embrittlement model: a model combining density functional theory data and transmutation helium production rates has been developed, which is able to predict critical helium grain boundary concentrations giving rise to helium embrittlement.
7. Integrated neutronics-microstructural evolution model for materials in a fusion power plant: a comprehensive model, combining simulation of neutron fields for an assumed detailed engineering model of a fusion power plant with a helium embrittlement model, has been developed and applied to assess the lifetime of various materials in a fusion power plant environment with respect to the onset of transmutation-induced helium embrittlement
8. Model for irradiation microstructure evolution observed in in-situ TEM experiments: a model has been developed for evolution of irradiation induced microstructure, including the statistics of defect production derived directly from cascade simulations and elastic interaction between defects formed in cascades. It has been applied to the interpretation of in-situ TEM experiments. For the first time the model explained the evolution of defect densities and size distributions in agreement with observations.
9. Multi-scale Finite Element Microstructure MEshfree fracture model (FEMME) for quasi-brittle materials with complex microstructures: a novel method using Cellular Automata integrated with Finite Elements has been developed that accounts for the effect of microstructure on quasi-brittle and thermal transport properties within finite element simulations of mechanical damage. The microstructure is modelled explicitly, and the computational efficiency is two orders faster than direct FE simulation at the same level of microstructural detail. This allows microstructure to be introduced into simulations of structural components, such as moderator bricks and fuel cladding. Validation experiments 3D studies of damage in polygranular graphite and SiC - SiCfibre ceramic composites.
Exploitation Route In using the new experimental and modelling methods we have developed in the characterisation of nuclear materials and the effects of radiiation damage on them.
Sectors Energy,Transport

Description 1) the sub-miniature testing techniques developed in the project formed a major part of an investigation into the causes of a jet-turbine failure for Rolls Royce. This was reported as an Impact Case Study in the Oxford Materials Department's recent REF submission. 2) the testing and analysis techniques developed in this project, and the partnerships established within and outside the UK, have been a very significant factor in Oxford and CCFE's involvement in setting up and operating the National Nuclear User Facility, not least the new-build on the CCFE site of the open-access Materials Research Facility. This facility now routinely uses methods developed in Oxford to study real nuclear materials for use in both fission and fusion. 3) methods developed as part of this grant for analysis of irradiated atom probe samples has led to further work on irradiated materials in both the UK and EU
First Year Of Impact 2013
Sector Energy,Transport
Impact Types Economic

Description EURATOM/UKAEA Fusion Association 
Organisation European Atomic Energy Commission (EURATOM)
Country Belgium 
Sector Public 
PI Contribution Experimental research on effects of radiation on materials microstructure and properties.
Collaborator Contribution Modelling of defects produced by radiation damage and their effects on materials microstructure and properties.
Impact See other sections.