There are 6 prospective Generation-IV nuclear reactor conceptual designs. SuperCritical Water-cooled nuclear Reactors (SCWRs) are one of these design options. The reactor coolant in SCWRs will be light water operating at 25 MPa and up to 625°C, actually at conditions above the critical point of water (22.1 MPa and 374°C, respectively). Current Nuclear Power Plants (NPPs) around the world operate at sub-critical pressures and temperatures achieving thermal efficiencies within the range of 30–35%. One of the major advantages of SCWRs is increased thermal efficiency up to 45–50% by utilizing the elevated temperatures and pressures. SuperCritical Water (SCW) behaves as a single-phase fluid. This prevents the occurrence of “dryout” phenomena. Additionally, operating at SCW conditions allows for a direct cycle to be utilized, thus simplifying the steam-flow circuit. The components required for steam generation and drying can be eliminated. Also, SCWRs have the ability to support hydrogen co-generation through thermochemical cycles. There are two main types of SCWR concepts being investigated, Pressure-Vessel (PV) and Pressure-Tube (PT) or Pressure-Channel (PCh) reactors. The current study models a single fuel channel from a 1200-MWel generic PT-type reactor with a pressure of 25 MPa, an inlet temperature of 350°C and an outlet temperature of 625°C. Since, SCWRs are presently in the design phase there are many efforts in determining fuel and sheath combinations suited for SCWRs. The design criterion to determine feasible material combinations is restricted by the following constraints: 1) The industry accepted limit for fuel centreline temperature is 1850°C, and 2) sheath-material-temperature design limit is 850°C. The primary candidate fuel is uranium dioxide. However; previous studies have shown that the fuel centreline temperature of an UO2 pellet might exceed the industry accepted limit for the fuel centreline temperature. Therefore, investigation on alternative fuels with higher thermal conductivities is required to respect the fuel centreline temperature limit. Sheath (clad) materials must be able to withstand the aggressive SCW conditions. Ideal sheath properties are a high-corrosion resistance and high-temperature mechanical strength. Uranium dicarbide (UC2) is selected as a choice fuel, because of its high thermal conductivity compared to that of conventional nuclear fuels such as UO2, Mixed OXide (MOX) and Thoria (ThO2). The chosen sheath material is Inconel-600. This Ni-based alloy has high-yield strength and maintains its integrity beyond the design limit of 850°C. This paper utilizes a generic SCWR fuel channel containing a continuous 43-element bundle string. The bulk-fluid, sheath and fuel-centreline temperature profiles together with Heat Transfer Coefficient (HTC) profile were calculated along the heated length of a fuel channel at the maximum Axial Heat Flux Profiles (AHFPs).
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18th International Conference on Nuclear Engineering
May 17–21, 2010
Xi’an, China
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-4930-9
PROCEEDINGS PAPER
Thermal Aspects of Using Uranium Dicarbide Fuel in an SCWR at Maximum Heat-Flux Conditions
Caleb Pascoe
,
Caleb Pascoe
University of Ontario Institute of Technology, Oshawa, ON, Canada
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Ashley Milner
,
Ashley Milner
University of Ontario Institute of Technology, Oshawa, ON, Canada
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Hemal Patel
,
Hemal Patel
University of Ontario Institute of Technology, Oshawa, ON, Canada
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Wargha Peiman
,
Wargha Peiman
University of Ontario Institute of Technology, Oshawa, ON, Canada
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Graham Richards
,
Graham Richards
University of Ontario Institute of Technology, Oshawa, ON, Canada
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Lisa Grande
,
Lisa Grande
University of Ontario Institute of Technology, Oshawa, ON, Canada
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Igor Pioro
Igor Pioro
University of Ontario Institute of Technology, Oshawa, ON, Canada
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Caleb Pascoe
University of Ontario Institute of Technology, Oshawa, ON, Canada
Ashley Milner
University of Ontario Institute of Technology, Oshawa, ON, Canada
Hemal Patel
University of Ontario Institute of Technology, Oshawa, ON, Canada
Wargha Peiman
University of Ontario Institute of Technology, Oshawa, ON, Canada
Graham Richards
University of Ontario Institute of Technology, Oshawa, ON, Canada
Lisa Grande
University of Ontario Institute of Technology, Oshawa, ON, Canada
Igor Pioro
University of Ontario Institute of Technology, Oshawa, ON, Canada
Paper No:
ICONE18-29974, pp. 767-774; 8 pages
Published Online:
April 8, 2011
Citation
Pascoe, C, Milner, A, Patel, H, Peiman, W, Richards, G, Grande, L, & Pioro, I. "Thermal Aspects of Using Uranium Dicarbide Fuel in an SCWR at Maximum Heat-Flux Conditions." Proceedings of the 18th International Conference on Nuclear Engineering. 18th International Conference on Nuclear Engineering: Volume 2. Xi’an, China. May 17–21, 2010. pp. 767-774. ASME. https://doi.org/10.1115/ICONE18-29974
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