Current CANDU-type nuclear reactors use a once-through fuel-channel with an annulus gas insulating it from the moderator. The current reference design for a CANDU-type SuperCritical Water-Cooled Reactor (SCWR) is to eliminate the annulus gap and use a ceramic insert to insulate the coolant from the moderator. While such a design may work, alternative fuel-channel design concepts are under development to explore the optimum efficiency of SCWRs. One such alternative approach is called the Re-Entrant fuel-channel. The Re-Entrant fuel-channel consists of three tubes, the inner tube (flow tube), pressure tube and an outer tube. The fuel bundles are placed in the inner tube. An annulus is formed between the flow and pressure tubes, through which the primary coolant flows. A ceramic insulator is placed between the pressure tube and the outer tube. The coolant flows through the annulus receiving heat from the inner tube from one end of the channel to another. At the far end, the flow will reverse direction and enter the inner tube, and hence the fuel-string. At the inlet, the temperature is 350°C for a high-pressure coolant (pressure of 25 MPa), which is just below the pseudocritical point. At the outlet, the temperature is about 625ºC at the same pressure (the pressure drop is small and can be neglected). The objective of this work was to design the Re-Entrant channel and to estimate the heat loss to the moderator for the proposed new fuel-channel design. A numerical model was developed and MATLAB was used to calculate the heat loss from the insulated Re-Entrant fuel-channel along with the temperature profiles and the heat transfer coefficients for a given set of flow, pressure, temperature and power boundary conditions. Thermophysical properties were obtained from NIST REFPROP software. With the results from the numerical model, the design of the Re-Entrant fuelchannel was optimized to improve its efficiency / UOIT

Under normal operating conditions a CANDU reactor pressure tube (PT) is insulated from its outer calandria tube (CT) by a CO2 gas annulus. If the primary loop coolant flow is compromised the PT can overheat and, if still pressurized, balloon into contact with the CT. At this point the moderator acts as an emergency heat sink. If the heat transferred from the CT to the moderator exceeds the critical heat flux (CHF) the CT can overheat, begin to strain due to the contact pressure, and eventually fail. A mechanistic model is presented that describes ballooning contact of the PT and CT, the resulting thermal contact conductance, heat flux to the moderator, and, if CHF is exceeded, the development of film boiling and potential CT strain. The goal is to create a software package that predicts fuel channel failure during a pressure tube overheat event. / Thesis / Master of Applied Science (MASc) / Computer software was developed to predict CANDU fuel channel failure in the event of a total station blackout. The model created successfully predicted the available experimental data.

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. Since fuel bundles are continuously under high neutron fluxes, annulus gas nuclides undergo different nuclear transformations generating new composition of the gas that might have different physical properties which are undesirable for the annulus system. In addition, gas nuclides become radioactive and lead to an increase of the radioactive material inventory in the reactor and consequently to an increase of radiation levels. Pressure Tube Reactor (PTR) and Pressure Tube Supercritical Water Reactor (PT SCWR) fuel channel models have been developed in Monte Carlo N-Particle (MCNP) code. Neutron fluxes in the fuel channel annulus gas have been obtained by simulating different types of neutron sources in both PTR and PT SCWR fuel channels. Transmutation rates of annulus gases have been calculated for different gases (CO2, N2, Ar and Kr) at different pressures and temperatures in both fuel channels. The variation of the transmutation rates, neutron fluxes and gas densities in the annulus gas have been investigated in PTR and PT SCWR fuel channels at constant pressures and different temperatures. MCNP code along with NIST REFPROP [14] and other software tools have been used to conduct the calculations. / UOIT

Fuel channels are used to keep the components of a nuclear fuel bundle in place. The fuel channel of a new nuclear fuel which is being developed at Westinghouse has a complex geometry which creates challenges in the manufacturing process. FE simulations were developed of a two-stage forming process of the fuel channel. Three different simulations were developed, a simulation of the pre-bending, a simplified simulation of bending, and a combined simulation of the two-stage process of pre-bending and bending. The simulations were done in Ansys Workbench. The simulations of the pre-bending could be validated against experimental results. The simulations of the bending showed significant differences in results compared to experiments. A few different sources of error were investigated due to the difference in results. / Höljerör omsluter kärnbränslet i en kärnreaktor och håller komponenterna i bränslepaketet på plats. Höljeröret av ett nytt kärnbränsle som utvecklas på Westinghouse har en komplex geometri som skapar utmaningar i tillverkningsprocessen. FE simuleringar av en tvåstegsprocess av bockningen av höljeröret utvecklades. Tre typer av simuleringar utvecklades, en simulering av förbockningen, en förenklad simulering av bockningen, och en simulering av den kombinerade förbockningen och bockning. Simuleringarna gjordes i Ansys Workbench. Förbockningssimuleringarna kunde valideras mot experiment. Bockningssimuleringen visade dock signifikanta skillnader i resultat gentemot experiment. Några olika felkällor undersöktes på grund av skillnaderna i resultat.