Master Theses & Projects (FESNS)

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Browsing Master Theses & Projects (FESNS) by Subject "Alternative fuels"

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    Thermal aspects of high efficiency channel with conventional and alternative fuels in SuperCritical water-cooled reactor (SCWR) applications
    (2011-03-01) Peiman, Wargha; Pioro, Igor; Gabriel, Kamiel
    Chosen as one of six Generation‒IV nuclear-reactor concepts, SuperCritical Water-cooled Reactors (SCWRs) are expected to have high thermal efficiencies within the range of 45 ‒ 50% owing to reactor‘s high outlet temperatures. A generic pressure-channel (or pressure-tube)SCWR operates at a pressure of 25 MPa with inlet- and outlet-coolant temperatures of 350°C and 625°C. Consequently, the sheath and fuel centerline temperatures are higher in SCWRs than those of the current nuclear reactors. Previous studies have shown that the sheath and fuel centerline temperatures could exceed the design and industry accepted limits of 850°C and 1850°C, respectively. These studies correspond to UO2 enclosed in a 43-element fuel bundle at an average thermal power per channel of 8.5 MWth. Additionally, these high operating conditions in the range of 350 - 625°C lead to high heat losses from the coolant to the moderator, which in turn reduces the overall thermal efficiency of the Nuclear Power Plant (NPP). Therefore, there is a need for alternative fuels or fuel bundles for future use in SCWRs. Hence, it is also necessary to determine the amount of heat losses from a number of fuel-channel designs for SCWRs. The objectives of this study are to investigate the possibility of using alternative fuels and to determine the heat losses from a fuel-channel design at SCWR conditions. The investigated fuels are categorized as low thermal-conductivity (e.g., UO2, MOX, and ThO2), high thermal-conductivity (e.g., UC, UC2, UN), and enhanced thermal-conductivity (e.g., UO2‒SiC, UO2‒C, and UO2‒BeO) fuels. Additionally, the examined fuel channel is the High Efficiency Channel (HEC), which has been designed by the Atomic Energy of Canada Limited (AECL) for the proposed CANDU SCWR. In order to achieve the objectives of this study, a steady-state one-dimensional heat-transfer analysis was conducted. The MATLAB© and NIST REFPROP© software were used for programming and retrieving thermophysical properties of a light-water coolant, respectively. The fuel centerline temperature was calculated for the fuel channels with the maximum thermal power, i.e., +15% above average channel power. Results of this analysis showed that the fuel centerline temperatures of low thermal-conductivity fuels exceed the industry limit; therefore, either a fuel with a higher thermal conductivity should be used or the fuel bundle geometry must be modified. Among the high thermal-conductivity fuels, UC has been shown to be a candidate for future use in SCWRs. However, the chemical compatibility of UC with water at high operating temperatures of SCWRs remains ambiguous. Therefore, further studies are required before selecting UC. In regards to enhanced thermal-conductivity fuels, UO2‒BeO is the most suitable candidate; however, its mechanical and neutronic properties must be thoroughly studied before any decision is made with regards to the selection of a fuel. In regards to the heat losses from the examined fuel channel, the heat loss was between 70 kW and 110 kW per fuel channel based on an average thermal power per channel of 8.5 MWth and a moderator pressure of 0.1 MPa at 80°C. A sensitivity analysis of the fuel channel shows that the heat loss can be reduced by increasing the operating pressure of the moderator, which in turn allows for increasing the operating temperature of the moderator. Higher operating temperatures of the moderator result in smaller temperature differences between the coolant and the moderator, which leads to lower heat losses. Therefore, either the thickness of the insulator or the pressure of the moderator should be increased in order to reduce the heat losses from the fuel channel.
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    Thermal aspects of using alternative nuclear fuels in supercritical water-cooled reactors
    (2010-11-01) Grande, Lisa Christine; Pioro, Igor
    A SuperCritical Water-cooled Nuclear Reactor (SCWR) is a Generation IV concept currently being developed worldwide. Unique to this reactor type is the use of light-water coolant above its critical point. The current research presents a thermal-hydraulic analysis of a single fuel channel within a Pressure Tube (PT) - type SCWR with a single-reheat cycle. Since this reactor is in its early design phase many fuel-channel components are being investigated in various combinations. Analysis inputs are: steam cycle, Axial Heat Flux Profile (AHFP), fuel-bundle geometry, and thermophysical properties of reactor coolant, fuel sheath and fuel. Uniform and non-uniform AHFPs for average channel power were applied to a variety of alternative fuels (mixed oxide, thorium dioxide, uranium dicarbide, uranium nitride and uranium carbide) enclosed in an Inconel-600 43-element bundle. The results depict bulk-fluid, outer-sheath and fuel-centreline temperature profiles together with the Heat Transfer Coefficient (HTC) profiles along the heated length of fuel channel. The objective is to identify the best options in terms of fuel, sheath material and AHFPS in which the outer-sheath and fuel-centreline temperatures will be below the accepted temperature limits of 850°C and 1850°C respectively. The 43-element Inconel-600 fuel bundle is suitable for SCWR use as the sheath-temperature design limit of 850°C was maintained for all analyzed cases at average channel power. Thoria, UC2, UN and UC fuels for all AHFPs are acceptable since the maximum fuel-centreline temperature does not exceed the industry accepted limit of 1850°C. Conversely, the fuel-centreline temperature limit was exceeded for MOX at all AHFPs, and UO2 for both cosine and downstream-skewed cosine AHFPs. Therefore, fuel-bundle modifications are required for UO2 and MOX to be feasible nuclear fuels for SCWRs.