Molten salt technologies are likely to be key components in future nuclear fuel cycles [1-3] and are of interest as coolants in fusion reactors as well as energy storage media to support solar power generation. The attraction of these technologies can be argued is predominantly based upon the inherent safety imparted by the use of molten salt media (e.g. reduced criticality risk allowing small plant footprints; reduced likelihood of volatile radioactive species formation compared to oxide fuel reactors; plant operation at atmospheric pressures; negligible radiation degradation of molten salt media [4-5]). Despite the advantages that molten salt technologies can bring to nuclear and non-nuclear processes, there are still considerable technical challenges that need to be overcome in order for deployment to be approved in current regulatory regimes. One of these major challenges is understanding the potential tritium behaviour in molten media, both in terms of removal and recovery for future deployment in fusion systems, and detritiation of the molten salt for end of life decommissioning. Tritium, a radioactive isotope of hydrogen, is continuously produced in molten salts such as FLiBe in nuclear reactor environments (both fission and fusion), therefore the transport and chemical behaviour of tritium needs to be well-characterised for the design of tritium control systems and for safety analysis.
Further challenges include understanding materials compatibility with molten salts, in terms of behaviour and performance of materials, that are in contact with molten salts which have tritium present due the extreme environments and potential to induce corrosion/degradation, leading to the formation of undesirable compounds (e.g., TF, the tritium isotopologue of hydrofluoric acid), and equipment design. Waste management and decommissioning of molten salts have not received much attention, but are key issues to determine their feasibility in fusion power plants. Therefore, removal of tritium from the molten material is a critical processing stage in any fusion fuel cycle deploying a molten breeder. At the end of life, it will also be a critical part of decommissioning to reduce the tritium content of the molten breeder material as much as possible. Although these two different tritium removal requirements are likely to require different hardware solutions, there are anticipated to be many synergies in the fundamental understanding of tritium extraction from the materials.
FLIBe is a promising molten salt candidate for future deployment as is possesses a volumetric heat capacity similar to water and the neutronic and thermohydraulic properties of FLiBe, and similar molten salts, make them strong candidates to be used as a coolant and breeder or only as the breeder material in fusion and further fission systems [1, 2]. However, the stability of this salt is critical to reactor safety and its continuous production of tritium, meaning it is vital to understand the chemical behaviour of the salts under these conditions with the control of tritium transport.
The overall project hypothesis is to ascertain the chemical and transport behaviour of tritium and further hydrogen isotopes in molten salt systems (FLiBe and others) in order to improve tritium management in nuclear fission and fusion reactors.
The project aims and objectives are consequently as follows:
· Key aim 1: Development of techniques that can readily quantify the chemical form of tritium in various molten salt media and its stability under irradiation.
o Objective 1a: Develop methods that can identify and quantify tritium speciation and transport in various molten salt media of interest.
o Objective 1b: Determine the viability of electrochemical probes that can provide information on interfacial behaviour in situ.
o Objective 1c: Assess the impact of relevant irradiation fields on the salt/materials interface in the presence of hydrogen isotopes for selected materials.
· Key aim 2: Develop methodologies for tritium/ other hydrogen isotopes removal and its recovery from molten salt media to allow the implementation of advanced molten salts technologies across the nuclear industry.
o Objective 2a: Assess how tritium removal affects the degradation mechanisms of selected molten salts/ media impacts on process performance for identified molten salt technologies
o Objective 2b Assess the effects of pure, dry salt mixtures on the type and extent of any degradation observed.
o Objective 2c: Quantify the recovery rates and purity of extracted tritium
· Key aim 3: Implementation and integration for future fusion systems
o Objective 3a: Develop an understanding of the requirements for scale up and integration with the UKAEA tritium extraction system for use in the fuel cycle of a fusion reactor using a molten salt blanket
This project will utilise state-of-the art nuclear Nuclear Materials facilities at the Henry Royce Institute (https://www.royce.ac.uk/research-areas/nuclear-materials/), the Dalton Cumbrian Facility (https://www.dalton.manchester.ac.uk/research/facilities/cumbria-facilities) and the Molten Salts in Nuclear Technology Laboratory (MSNTL) National Nuclear User Facility (https://www.nnuf.ac.uk/molten-salts-nuclear-technology-laboratory). The MSNTL aims to be capable of handling most fluoride salts by early 2023 with further developments in expertise and capability to allow the handling of FLiBe by 2024. The student will have the opportunity to spend time at UKAEA Hydrogen-3 Advanced Technology (H3AT) facility to develop an understanding of the requirements for a tritium extraction system. Finally the project will involve secondments at the USA National laboratory Oak Ridge national laboratory to develop skills and experience in advanced nuclear technologies.
This position is only available to UK students.
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