Skip to main content
UOIT Home
Please use this identifier to cite or link to this item: http://hdl.handle.net/10155/827

Issue Date: 1-Aug-2017
Title: Study on specifics of thermalhydraulics and neutronics of pressure-channel supercritical water-cooled reactors (SCWRs)
Authors: Peiman, Wargha
Publisher : UOIT
Degree : Doctor of Philosophy (PhD)
Department : Nuclear Engineering
Supervisor : Pioro, Igor
Gabriel, Kamiel
Keywords: SCWR
Thermalhydraulics
Neutronics
Generation IV
Nuclear fuel
Abstract: 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.
Appears in Collections:Electronic Theses and Dissertations (Public)
Faculty of Energy Systems and Nuclear Science - Doctoral Dissertations

Files in This Item:

File Description SizeFormat
Peiman_Wargha.pdf9.68 MBAdobe PDFView/Open

Items in e-scholar@UOIT are protected by copyright, with all rights reserved, unless otherwise indicated.