Corrosion and environmentally-assisted cracking research
Understanding the corrosion mechanisms of materials in extreme environments is providing the drive to establish new approaches to predict the relationship between environment, microstructure and stress.
Corrosion of ILW Containers
Elsie Onumonu, PhD Student
Intermediate Level Nuclear Waste, (which can be metallic, oxides or mixtures of types of materials) is encapsulated in cement and stored in 500 litre 316L Stainless Steel drums, in various forms defined by NIREX.
The cement environment is normally protective towards steel of all types due to the high internal pH, but aggressive ions can promote local corrosion on the interior of the drum. The effects of the active waste in emitting both radiation and heat can increase the corrosivity of the aqueous solution within the cemented wasteform. The drum is expected to maintain its mechanical strength and prevent the release of material for a period of at least 500 years.
Defects in Proton-Irradiated Materials
Jonathan Duff, PhD Student
Neutron irradiated stainless steels become susceptible to irradiation assisted stress corrosion cracking (IASCC) in oxygenated high temperature water environments. Intergranular cracking in metals is strongly influenced by grain boundary character. However, the effect of boundary character on IASCC is unknown. Both segregation and irradiation hardening are considered contributing factors to IASCC and may depend on boundary structure. Further details on Defects in Proton-Irradiated Materials.
IASCC and Atmospheric Corrosion
Tony Cook, PDRA
This project is part of the EPSRC Keeping the Nuclear Option Open programme. The research programme will establish new expertise alongside new mechanistic understanding and predictive models for the assessment of materials performance in existing/potential new plant and for intermediate level waste (ILW) storage. Fundamental research into two life-limiting environmental degradation mechanisms will be addressed by experiment, analysis and modelling: (i) irradiation-assisted stress corrosion cracking (IA-SCC) of core internal materials in high temperature water, and (ii) atmospheric-induced stress corrosion cracking (AI-SCC) of ILW container material under intermittent chloride supply.
Investigative Contamination Experiments Related to Steel Storage Vessels in the Nuclear Power Industry
Guy Woodhouse, PhD Student
This project focuses on the development and application of practical decontamination technologies for steel and other structures contaminated with radioactive materials. The general aim of this project is to explore alternatives to "traditional" approaches to metal decontamination, such as the use of mineral acid mixtures or similar reagents, which generate large volumes of chemically hazardous, secondary radioactive wastes which themselves require treatment. Further details on Investigative Contamination Experiments Related to Steel Storage Vessels in the Nuclear Power Industry.
Kinetic Studies in Cement
Evripidis Tsaousoglou, PhD Student
The electrical properties of cement and encapsulated metal waste systems evolve as the cement sets and may evolve further as the metal surface corrodes and reacts with the cement in the case of alkali metals. Electrochemical testing is a simple way to monitor these processes. Further details on Kinetic Studies in Cement.
Magnox Pond Corrosion
Rob Burrows, PhD Student
The integrity of spent fuel element clad is imperative in avoiding undesirable release of soluble fission products into cooling ponds following discharge from reactor into aqueous storage. Corrosion is a possible route to clad penetration and pond chemistry is controlled to maintain passivity. The aqueous corrosion behaviours of commercial magnesium alloy Magnox Al80 and natural uranium have been studied with the use of modern electrochemical techniques
Mechanical Modelling of Intergranular Stress Corrosion Cracking
Andrey Jivkov, PDRA
Stress corrosion cracking presents a considerable hazard to both safety and economic performance in many industrial sectors, because the lifetime prediction is influenced by uncertainties in short crack behaviour. One source of the uncertainty is the material’s microstructure, which determines the resistance to intergranular stress corrosion cracking. For example, the random grain boundaries, which are prone to sensitisation, form paths of low resistance for intergranular cracks to follow. The non-sensitised special grain boundaries are observed to encourage crack bridging ligament formation. In this project, 2D and 3D finite element models for intergranular cracking have been developed, with the aim of a quantitative investigation of crack bridging and its effects on crack propagation. Further details on Mechanical Modelling of Intergranular Stress Corrosion Cracking.
Mechanism of the Immunity of Alloy 690 to Primary-Water Stress Corrosion Cracking
Fabio Scenini, PhD Student
PWSCC of Alloy 600 is now a very well-known phenomenon, and replacement or substitution by the high-Cr Alloy 690 is well under way worldwide. Recently evidence has been growing that the superficial oxidation reactions of the alloy control the occurrence of SCC; in particular, electropolished surfaces of Alloy 600 show internal oxidation of Cr at low potentials or in hydrogen/steam environments where the potential is below Ni/NiO; Alloy 690 shows external oxidation, which is one proposed reason for its greater resistance to SCC. The situation for mechanically prepared surfaces is more complex, and for a crack tip more complex still. A better understanding of oxide development on these materials is crucial to the validation of the long-term performance of Alloy 690. Further details on Mechanism of the Immunity of Alloy 690 to Primary-Water Stress Corrosion Cracking.
Stress Corrosion Cracking in Boiling and Pressurised Water Reactors
Kuvasani Govender, PDRA
Stress corrosion cracking (SCC) is the term given to the phenomenon that occurs under the synergistic and simultaneous operation of sustained tensile and chemical stresses on a susceptible material. In the context of nuclear power generation, efforts have been made to understand and alleviate SCC in austenistic stainless steels employed in boiling (BWRs) and pressurised water reactors (PWRs). Sensitized materials are known to be susceptible to SCC in oxygenated water found in BWRs. By contrast, in PWRs, which nominally operate under oxygen-free conditions, SCC problems are usually associated with stagnant conditions in occluded superheated crevices or "dead-space" regions, which are not easily deoxygenated at the plant initiation stage. Further details on Stress Corrosion Cracking in Boiling and Pressurised Water Reactors
Stress Corrosion Cracking of Austenitic Stainless Steels
Karen Shapiro, PhD Student
A combination of tensile stresses and a corrosive environment is one of the most important causes of failures of metal structures. This kind of attack is known as 'stress corrosion cracking' (SCC). It is defined as the spontaneous failure of metals as a result of the combined action of a corrosive environment and tensile stresses; either applied or residual. Further details on Stress Corrosion Cracking of Austenitic Stainless Steels.