We are thankful to be welcome on these lands in friendship. The lands we are situated on are covered by the Williams Treaties and are the traditional territory of the Mississaugas, a branch of the greater Anishinaabeg Nation, including Algonquin, Ojibway, Odawa and Pottawatomi. These lands remain home to many Indigenous nations and peoples.

We acknowledge this land out of respect for the Indigenous nations who have cared for Turtle Island, also called North America, from before the arrival of settler peoples until this day. Most importantly, we acknowledge that the history of these lands has been tainted by poor treatment and a lack of friendship with the First Nations who call them home.

This history is something we are all affected by because we are all treaty people in Canada. We all have a shared history to reflect on, and each of us is affected by this history in different ways. Our past defines our present, but if we move forward as friends and allies, then it does not have to define our future.

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march 2, 2016

Speaker: Ms. Ashlea Colton
Title: Improved full-core modelling of natural uranium fueled nuclear reactors through transport corrected diffusion coefficients
Abstract: Nuclear reactors are an integral component of Ontario's power grid, generating over 40% of the province's power needs. The heart of the nuclear reactor is the reactor core, which contains fissile material that can sustain a nuclear chain reaction. To safely operate a nuclear reactor, it is imperative to understand the behavior of the neutrons. The neutron distribution is described by the flux, a parameter defined as the number of neutrons at a given position, trajectory and energy. Two equations can be used to model the neutron flux: the neutron transport equation, and the neutron diffusion equation. While the transport equation is accurate, it is a computational challenge to evaluate with large geometries. It is therefore impractical for nuclear operators or reactor physicists to use the transport equation for full-core neutron flux predictions. The diffusion equation is used to determine the neutron flux in full core geometries quickly, but has its own challenges. The diffusion estimated flux values are inaccurate near void boundary conditions as would occur at the reactor boundary. This seminar describes a method to improve the accuracy of the neutron diffusion equation by deriving the diffusion coefficient using the transport calculated neutron flux for a one dimensional geometry. Comparisons are made between transport and diffusion fluxes for a number of geometries to evaluate the improvements achieved with the new diffusion coefficient. The results of these comparisons show promise as diffusion estimated neutron fluxes are closer to transport fluxes, with greatest gains achieved near the void boundary condition.