The program investigates CANDU pressure tube (PT) hydride blister formation that can lead to cracking. In-reactor gap monitoring is performed from within the PT by an eddy current (EC) probe. EC based gap measurement is used in predictions of time-to-contact between PT and calandria tube (CT) and is important for safety and licensing of CANDU reactors.

The project seeks to generate a comprehensive model of EC measurement of PT to CT gap beginning with a two-dimensional analytical model followed by three-dimensional FEM (COMSOL) modelling to quantify the effects of essential parameters on gap measurement accuracy.

As well, the project team will assemble an experimental set-up for laboratory PT to CT gap measurement using actual transmit-receive eddy current probe technology.

Ultimately, this research aims to improve gap measurement accuracy and make recommendations for achieving improvements in accuracy within existing gap measurement systems, with the goal of providing support for inspection qualification programs.

This research uses computational fluid dynamics (CFD) to obtain detailed information on the natural convection fluid flow and local and time-varying heat transfer rates at the vessel wall within CANDU reactors. This information will be used to assess the appropriateness of the underlying assumptions in the simplified integral codes which are currently used in the nuclear industry for beyond design basis analysis.

Under severe accident conditions, like what occurred at the Fukushima Daiichi nuclear plant in 2011, melting of the nuclear fuel rods and the interior reactor core components can occur forming a molten corium pool and leading, potentially, to a breach of the reactor vessel.

There is a need to ensure cooling on the exterior of the vessel is sufficient to maintain vessel integrity. The completed research will be used to assess the validity of the underlying assumptions in codes such as MAAP-CANDU which are used for assessment of in-vessel retention under severe accident conditions.

This research uses computational fluid dynamics (CFD) to obtain detailed information on the natural convection fluid flow and local and time-varying heat transfer rates at the vessel wall within CANDU reactors. This information will be used to assess the appropriateness of the underlying assumptions in the simplified integral codes which are currently used in the nuclear industry for beyond design basis analysis.

Under severe accident conditions, like what occurred at the Fukushima Daiichi nuclear plant in 2011, melting of the nuclear fuel rods and the interior reactor core components can occur forming a molten corium pool and leading, potentially, to a breach of the reactor vessel.

There is a need to ensure cooling on the exterior of the vessel is sufficient to maintain vessel integrity. The completed research will be used to assess the validity of the underlying assumptions in codes such as MAAP-CANDU which are used for assessment of in-vessel retention under severe accident conditions.

This research uses computational fluid dynamics (CFD) to obtain detailed information on the natural convection fluid flow and local and time-varying heat transfer rates at the vessel wall within CANDU reactors. This information will be used to assess the appropriateness of the underlying assumptions in the simplified integral codes which are currently used in the nuclear industry for beyond design basis analysis.

Under severe accident conditions, like what occurred at the Fukushima Daiichi nuclear plant in 2011, melting of the nuclear fuel rods and the interior reactor core components can occur forming a molten corium pool and leading, potentially, to a breach of the reactor vessel.

There is a need to ensure cooling on the exterior of the vessel is sufficient to maintain vessel integrity. The completed research will be used to assess the validity of the underlying assumptions in codes such as MAAP-CANDU which are used for assessment of in-vessel retention under severe accident conditions.

This research uses computational fluid dynamics (CFD) to obtain detailed information on the natural convection fluid flow and local and time-varying heat transfer rates at the vessel wall within CANDU reactors. This information will be used to assess the appropriateness of the underlying assumptions in the simplified integral codes which are currently used in the nuclear industry for beyond design basis analysis.

Under severe accident conditions, like what occurred at the Fukushima Daiichi nuclear plant in 2011, melting of the nuclear fuel rods and the interior reactor core components can occur forming a molten corium pool and leading, potentially, to a breach of the reactor vessel.

There is a need to ensure cooling on the exterior of the vessel is sufficient to maintain vessel integrity. The completed research will be used to assess the validity of the underlying assumptions in codes such as MAAP-CANDU which are used for assessment of in-vessel retention under severe accident conditions.

In the past decade, there has been tremendous progress in developing unmanned aerial vehicles (UAVs), also called drones. While first products targeted consumers, solutions for industrial use cases, like inspection, monitoring, and surveillance, are the next frontier.

The first solutions on the market are promising but require an experienced pilot to navigate the UAV. The main goal of this proposal is to develop a fully automated UAV monitoring solution for industrial sites. The integration of a combination of localization sensors will localize the vehicle with high accuracy indoors and outdoors. To monitor large industrial sites efficiently, we propose a multi-UAV strategy.

The University of Toronto (UofT) has extensive knowledge in concrete integrity and ageing assessment. For this knowledge to translate to a UAV-based solution, a lightweight sensor suite will be researched and demonstrated in proof-of-concept experiments with the goal of achieving automated, remote concrete health assessment. Ontario Power Generation (OPG) is an important partner on this project and will provide direct input regarding the requirements for a UAV-based monitoring solution.

The proposed research will develop expertise in the testing of graphene oxide membranes for isotopic filtration of water. The long-term goal is to develop a system capable of reliably filtering tritium from heavy water.

A computational modelling study will support the research to determine the underlying mechanism for the filtration and, hence, guide a process of optimizing the membrane through functional additions. Different membrane types will be assessed and optimized, and in addition, there will be an approach to scaling up from the lab to a pilot plant.

Eventually, nuclear reactors must be decommissioned. The key outcome of this research will be the efficient separation of tritiated water from the decommissioned heavy water moderator. Efficient filtration can make safe storage much easier. It will provide separated tritium and heavy water for subsequent sale and position Canada to lead the global marketplace for decommissioning reactors emerging internationally over the next few decades.

The focus of this research project is on developing automation and robotic methods, specifically, environment characterization to aid in the deconstruction of nuclear power plants. The ultimate goal of this work is to reduce exposure to humans from decommissioning activities, which includes demolition, packaging and removal of debris and safe storage activities.

It is anticipated that the results of this study will significantly enhance the tools and the process of providing semantic model inputs regarding the facility environment to decommissioning robots, automated waste selection and packing optimization for low to intermediate waste disposal.

Nearly a third of the world’s nuclear plants are over 30 years old and a large number of these will be decommissioned in the coming decades, including the Ontario Power Generation’s (OPG) Pickering Nuclear. The project aims to develop efficient methods to create as-built building information models, populate those models and advance concepts and tools used to address decommissioning-related inventory assessment and selective disassembly for optimization of demolition, disassembly and packing.

This research focuses on three themes: Low Energy Radionuclide Monitoring, Waste Capture and Decontamination, Waste Segregation and Decommissioning Project Management.

The approach under study is the use of a Gas-ionization type detector modified for the detection of low energy emissions such as those associated with Tritium and Fe-55. Different methods for converting tritiated heavy water into a vapour form for direct injection into a GEM detector, thus reducing the need for a much higher gain, have been studied. The results suggest either electrolysis or atomization methods should be used to maximize the amount of tritium in the vapour environment.

As well, the research considers the use of electrostatic or electrohydrodynamic forces and plasma-based techniques for the ionization and capture of radioactive species. The technique has demonstrated the proof of principle that Iodine vapour generated during decommissioning activities would be captured by electrostatic methods.

The work is still mostly in the developmental phase and it is expected, the database for decontamination selection could be used by Ontario Power Generation as a preliminary tool.