Key Challenge 3
Environmental Degradation
Leads: Chris Grovenor (Experimental) and Mark Wenman (Modelling)
The development of new Zr-alloys with greater in-reactor resistance to oxidation and hydrogen- embrittlement will require improving our understanding of the key mechanisms that control metal reactivity, transport processes, and the influence of microstructure. Key challenges will be:
Designing new experiments with our international partners to determine if the effects of and neutron irradiation are synergistic or competitive;
Developing new modelling techniques to predict the mechanistic effects of irradiation damage;
Further development of the advanced characterisation tools pioneered by the team for the study of the mechanisms of corrosion and application of these techniques to reactor-formed oxides;
Using the knowledge developed to explore the degradation of zirconium alloys in the different environments found in fusion reactors;
Exploring behaviour of the interface of Cr-coated Zr-alloys for the development of accident-tolerant fuel (ATF) cladding.
Work Packages
WP 3.1: Corrosion of Irradiated Materials
The development of true understanding of this area requires detailed analysis of irradiation-induced microstructural damage, and especially the re-distribution of alloying elements from SPPs into the metal matrix, which is carried out in Key Challenge 1. These well-characterised alloys will then be available for corrosion measurements in a dedicated active autoclave in the Henry Royce Institute. Initial corrosion tests on H+ irradiated samples will be used to down-select the most appropriate neutron-irradiated samples for controlled corrosion in simulated PWR water for times up to 250 days, and comparison with un-irradiated control samples. Additionally, samples corroded under mechanical load will be used as an analogue to irradiation-induced growth experienced in-reactor (see KC1), and investigate its influence on corrosion. Synchrotron XRD will evaluate the macroscopic effect of irradiation on the oxides that form, while bulk EBSD analysis of the oxides will assess the impact of irradiation on the formation of oxide macrozones, recently found to be very influential for non-irradiated samples. We will then select samples for the more sophisticated measurements of the nanostructures that control the corrosion behaviour. These will include high-resolution oxide grain mapping using TEM-based (Astar) and SEM-based (Transmission Kikuchi Diffraction) analysis of oxide grain structure, texture and phase fraction (h-ZrO, t-ZrO2 and m-ZrO2)1 and on the same samples detailed porosity measurements . Atom Probe Tomography analysis will be used to measure distributions of alloying elements in both metal matrix (KC-1) and oxide (KC-3). The newly developed in-situ corrosion capability, with simultaneous low-dose rate proton irradiation, at the Dalton Cumbrian Facility (DCF) will evaluate the contribution of irradiation during corrosion, which will be of particular interest when applied to the altered microstructures of the neutron-irradiated material. Team members at Imperial have recently shown that significant advances in the understanding of corrosion of zirconium alloys can be made through combining electronic structure DFT modelling with the kinds of detailed experimental data described above. These successful correlations were limited to small defects (vacancies or small clusters) by the computational costs associated with cells of more than 100-200 atoms in DFT. Empirical potentials do not account for charge transfer and therefore cannot be used for understanding corrosion. Work at Imperial is currently developing density functional tight-binding models (DFTB) for the Zr-H system, but corrosion clearly involves oxygen and alloying elements such as Fe, Nb, Sn, Cr, Ni as well, and in this WP we will develop DFTB models to include all these elements. Expansion of this work to Caesium and Iodine is also required for modelling work on Pellet Cladding Interaction in KC-4. DFTB runs at 100-1000 times the speed of pure plane-wave DFT codes and allows 10x the number of atoms for the same cpu time, now enabling the study of dislocations and grain boundaries, and how elements segregate to them. We will directly compare the nanoscale segregation phenomena analysed in n-irradiated materials by Atom Probe and TEM studies with models at relevant timescales and explore the key mechanisms for diffusion and segregation behaviour.
WP 3.1 Key Deliverables
WP 3.2: Corrosion Under Gamma Irradiation
Gamma-radiation may play an important role in Zr alloys corrosion, not only by direct irradiation damage, but by changes induced in the water chemistry. One of the possible effects of gamma irradiation on corrosion can be correlated to the generation of active radicals, some of which may diffuse through the growing oxide more rapidly than normal cathodic reaction products. Another suggestion for the effect of gamma-radiation is that it encourages the formation of excess porosity that accelerates corrosion and H pickup. MIDAS will use autoclaves attached to Co-60 sources at the Dalton Cumbrian Facility to provide irradiation dose rates up to 1kGy/hr, in heavy water at temperatures up to 250C. Bare metal and pre-oxidised materials will be studied to measure oxidation rates under irradiation, changes in oxide nano-structure after gamma irradiation, and differences in the corrosion behaviour of control and n-damaged materials. Samples after irradiation (and controls) will be chosen for the same nanostructural analysis as in WP3.1. In addition, 3D NanoSIMS datasets will be used to reveal the traps and pathways of deuterium through the oxide layer, to give direct evidence for how the irradiation may alter the mechanisms of H diffusion. This WP will benefit from collaboration with our partners at Jacobs, who are running a programme of work studying the effect of gamma irradiation on Zircaloy corrosion using in-situ electrochemical impedance spectroscopy, and will share samples for characterisation. After we have the first experimental results, we will consider the feasibility of adding a modelling component to this area.
WP 3.2 Key Deliverable
A unique data set to allow us to answer the question By what mechanisms does gamma-irradiation in the absence of neutrons contribute to the corrosion performance of Zr alloys?
WP 3.3: Microstructural models for oxide growth
Despite several decades of careful analysis of the morphology of the oxide/metal interface, we still cannot explain why the characteristic surface undulations occur, i.e. whether it is stress-diffusion controlled or mechanical buckling, or both. We also do not understand the role of second phase particles (SPPs) in the fracture process or the stress induced in these relatively noble particles as the oxide grows past them. This is a critical gap in understanding, since it is well known that all the rate-controlling electrochemical processes during in-service corrosion happen within 50-200nm of this interface, and that local morphology controls the local corrosion rate. This WP will contain the development of a new model and collation of existing and new experimental data to both inform and validate this model. Previous models have been unable to properly account for the response of the M/O interface to stress (measurements confirm the oxide is compressively stressed to 1-3 GPa). The use of local formulation models, e.g. finite element, have previously tried to predict this but cannot correctly deal with the multiple-cracking phenomena observed experimentally, the stress relaxation this causes in the oxide, or the formation of the characteristic undulations in the interface. We will use a non-local method of peridynamics to study oxide formation, with anisotropic expansion and local cracking explicitly accounted for. These models will allow the study of the anisotropic oxide mechanical properties at the microscale, and predict the crack growth and influence of the SPPs on this process. In order to base the model firmly on realistic microstructures, these predictions will be compared to experimental data of micro- and nano-cracking and stress on a wide variety of alloys/stages of oxidation (burnup). Differences in interface morphology observed in WP 3.1 for irradiated samples will be investigated with respect to the changes in the SPPs, chemical distribution (KC-1) and local mechanical properties (KC-2), due to irradiation damage. Together, these will provide insight into the contribution of stress accumulation in the oxide to the differences in porosity for in- reactor samples compared to those exposed to autoclave testing.
WP 3.3 Key Deliverable
The first robust, predictive model of the development of morphology at the M/O interface during the corrosion of zirconium, including the influence of irradiation damage.
WP 3.4: Barrier layers in LOCA-resistant materials
Since Fukushima, there has been a dramatic increase of work on accident-tolerant fuel cladding, a good deal focussing on the application of high-temperature steam-resistant coatings to commercial Zr alloys. Cr-based coatings are amongst the most promising, with different coating chemistries and methods of deposition explored. The formation of intermetallic has been reported more recently, and with concerns that these reactions may compromise safe operation of the fuel cladding in- service. Should the temperature reach 1332C, the eutectic reaction will ensure rapid degradation of structural integrity, so interlayers that act as diffusion barriers are now being included between the Cr coating and base metal. Current UK work on LOCA-resistant (loss of coolant accident) Zr fuel cladding includes the EU H20202 project Il Trovatore on carbide, oxide and Fe-alloy coatings, in which both Manchester and Oxford are partners, a BEIS-funded project under the fuel development programme, led by NNL with work on Cr coatings made by Manchester Metropolitan University and characterised at Manchester, a Westinghouse-sponsored PhD project at Manchester on cold-sprayed coatings, and an EDF/ Westinghouse PhD project on Cr-coating behaviour during LOCA. This WP is planned to start in Y3, and will build on this current work in the partnership, with a focus on exploring refractory nitride, sacrificial, and amorphous diffusion barriers for high-temperature protection. The principles of barrier layer stability have been comprehensively explored by the microelectronics industry where complex multilayer structures have to survive temperature excursions during the processing of metallisation, and it is these principles we will use to design and test Cr/X/Zr alloy couples under LOCA conditions. We plan to send from previous collaborations and new work to KIT for corrosion tests under LOCA conditions. Proton irradiation at the Dalton Cumbrian Facility will be used to evaluate irradiation tolerance of these materials and the effect of damage on oxidation in PWR conditions.
WP 3.4 Key Deliverable
Selection of promising barrier layer materials to improve the performance of Cr-coated Zr alloys under LOCA conditions.
WP 3.5: Corrosion and Hydrogen pickup in Zr for fusion
There are a number of strategic reasons for reconsidering the potential of Zr alloys as blanket materials in fusion reactor, and in this WP, we will explore the corrosion performance of modern commercial Zr alloys (from KC-1) under the conditions specified by our partners at UKAEA, specifically corrosion resistance with water in a temperature range above PWR conditions, and irradiation with a very different energy spectrum, where He bubble formation (not usually observed LWR cladding) may occur. H-pickup is also a key design parameter that we have been encouraged to explore. Recent work has shown that He bubble formation in Zr alloys may have a significant effect on H trapping, and dual beam H/He irradiations (at DCF) will be used to investigate the importance of this effect on subsequent corrosion mechanisms. With CCFE, we will define a matrix of experiments (T, t, alloy choice), after consulting with our industrial nuclear fission partners on the results of work on accelerated temperature testing of Zr in steam. We will then carry out this chosen matrix of corrosion tests for up to 250 days on test coupons. These coupons will then be the sample set on which to carry-out the characterisation of key parameters using the advanced analytical techniques that the MIDAS team has deployed on materials corroded under fission conditions. This will include (with increasing levels of complexity) measurements of oxide growth kinetics, oxide macro- and micro-structure, oxide stress, texture and porosity, H-pickup fractions, H diffusion kinetics and penetration mechanisms (with Deuterium spiking).
WP 3.5 Key Deliverable
The first complete body of data on high-temperature corrosion processes and mechanisms for Zr alloys under conditions directly relevant to the fusion community.