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Development of a 37-element fuel bundle for the production of molybdenum-99 in CANDU power reactors.
(2014-08-01)
In this study, the potential use of CANDU power reactors for the production of Mo-99 is assessed. Five different modifications of a 37-element fuel bundle that could be used for the production of Mo-99 in existing CANDU ...
A Computational Fluid Dynamics based model that predicts wall shear stress in CANDU outlet feeder pipes
(2018-01-01)
Wall thinning of carbon steel in CANDU reactor outlet feeder pipes due to Flow Accelerated Corrosion (FAC) is identified as one of the challenges for CANDU reactors since it would force them to shut down due to safety ...
Petri net modeling of fault analysis for probabilistic risk assessment
(2013-04-01)
Fault trees and event trees have been widely accepted as the modeling strategy to perform Probabilistic Risk Assessment (PRA). However, there are several limitations associated with fault tree/event tree modeling. These ...
Transmutation rates in the annulus gas of pressure tube water reactors
(2011-07-01)
CANDU (CANada Deuterium Uranium) reactor utilizes Pressure Tube (PT) fuel channel design and heavy water as a coolant. Fuel channel annulus gas acts as an insulator to minimize heat losses from the coolant to the moderator. ...
Moderator displacers for reducing coolant void reactivity in CANDU reactors: a neutronics scoping study.
(2014-08-01)
When the coolant is voided in a CANDU lattice, the net reactivity change is positive, due primarily to the fact that the coolant and moderator are separated and the coolant volume is much smaller than the moderator volume. ...
Localization of a robotic crawler for CANDU fuel channel inspection
(2017-06-01)
This thesis discusses the design and development of a pipe crawling robot for the purpose of CANDU fuel channel inspection. The pipe crawling robot shall be capable of deploying the existing CIGAR (Channel Inspection and ...
Transport-theory-equivalent diffusion coefficients for node-homogenized neutron diffusion problems in CANDU lattices
(2010-04-01)
Calculation of the neutron flux in a nuclear reactor core is ideally performed by solving the neutron transport equation for a detailed-geometry model using several tens of energy groups. However, performing such detailed ...
Simulation of in-core dose rates for an offline CANDU reactor
(2016-04-01)
This thesis describes the development of a Monte Carlo simulation to predict the dose rates that will be encountered by a novel robotic inspection system for the pressure tubes of an offline CANDU reactor. Simulations were ...
Leakage-corrected discontinuity factors for a second-generation Th-Pu pressure-tube SCWR.
(2014-08-01)
The neutron flux throughout a reactor core should be calculated, ideally, by solving the neutron transport equation for a highly detailed geometric model of the core. Since this is computationally impractical, approximate ...
Design and development of a robotic crawler for CANDU fuel channel inspection
(2015-08-01)
For the design of a new robotic crawler drive unit for CANDU fuel channel inspection, a complete design and screening process was done in order to fulfil the objective of this research. A brief explanation of CANDU reactors ...