Browsing by Author "Peiman, Wargha"

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Study on specifics of thermalhydraulics and neutronics of pressure-channel supercritical water-cooled reactors (SCWRs)
(2017-08-01) Peiman, Wargha; Pioro, Igor; Gabriel, Kamiel
A group of countries has initiated an international collaboration to develop a next generation (i.e., Generation IV) of nuclear reactors. Chosen as one of the six Generation‒IV nuclear-reactor concepts, the SCWRs are expected to have high thermal efficiencies within the range of 40 ‒ 50% owing to reactor’s high outlet temperatures. The Canadian pressure-tube-type SCWR is featured with 3-batch refueling, 336 vertical fuel channels, a porous ceramic insulator inside the pressure tube, and stainless-steel cladding. The reactor operates at a pressure of 25 MPa with the coolant temperature rising from 350 to 625°C. Consequently, sheath and fuel centerline temperatures are significantly higher in SCWRs compared to those of the current water-cooled nuclear reactors. The main objective of this thesis is to conduct a study on specifics of the thermalhydraulics and neutronics of a pressure-tube SCWR based on an understanding of the supercritical water phenomena and their impacts on reactor design and operation. This thesis investigates the impact of several thermalhydraulic modeling parameters on fuel and cladding temperatures of a pressure-tube SCWR. The investigated thermalhydraulic modeling parameters are: 1) variable heat transfer coefficient, which is affected by thermophysical properties of supercritical water, axial heat flux, and three heat-transfer regimes: normal, improved and deteriorated; 2) thermophysical properties, which are affected by the bulk-fluid-temperature profile along the heated length and pressure drop along the fuel channel; 3) variable axial and radial heat-flux profiles of a fuel assembly (bundle string), which are affected by the neutron flux; 4) radial non-uniform heat generation inside the fuel; 5) axial and radial variable thermal conductivity of a fuel; 6) contact thermal resistance between the fuel and cladding; 7) heat loss from the coolant to the moderator, which is affected by the thermal conductivity of a ceramic insert; and 8) pressure drop of the coolant along the fuel channel. The main neutronic aspects, which have been incorporated in the neutronic model, include 1) variable coolant density along the heated length of the fuel channel, which affects neutronic properties of a lattice and, hence, the neutron flux and 2) number of energy groups, which affects the calculated channel powers.
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.