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Research Papers: Nuclear Power

# Thermal Aspects of Uranium Carbide and Uranium Dicarbide Fuels in Supercritical Water-Cooled Nuclear Reactors

[+] Author and Article Information
Lisa Grande

Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, Ontario, L1H 7K4, Canadalisa.grande@mycampus.uoit.ca

Bryan Villamere

Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, Ontario, L1H 7K4, Canadabryan.villamere@mycampus.uoit.ca

Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, Ontario, L1H 7K4, Canada

Igor Pioro

Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, Ontario, L1H 7K4, Canadaigor.pioro@uoit.ca

J. Eng. Gas Turbines Power 133(2), 022901 (Oct 25, 2010) (7 pages) doi:10.1115/1.4001299 History: Received January 18, 2010; Revised January 21, 2010; Published October 25, 2010; Online October 25, 2010

## Abstract

Supercritical water-cooled nuclear reactors (SCWRs) are a Generation IV reactor concept. SCWRs will use a light-water coolant at operating parameters set above the critical point of water (22.1 MPa and $374°C$). One reason for moving from current Nuclear Power Plant (NPP) designs to SCW NPP designs is to increase the thermal efficiency. The thermal efficiency of existing NPPs is between 30% and 35% compared with 45% and 50% of supercritical water (SCW) NPPs. Another benefit of SCWRs is the use of a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc. can be eliminated. Canada is in the process of conceptualizing a pressure tube (PT) type SCWR. This concept refers to a $1200-MWel$ PT-type reactor. Coolant operating parameters are as follows: a pressure of 25 MPa, a channel inlet temperature of $350°C$, and an outlet temperature of $625°C$. The sheath material and nuclear fuel must be able to withstand these extreme conditions. In general, the primary choice for the sheath is a zirconium alloy and the fuel is an enriched uranium dioxide $(UO2)$. The sheath-temperature design limit is $850°C$, and the industry accepted limit for the fuel centerline temperature is $1850°C$. Previous studies have shown that the maximum fuel centerline temperature of a $UO2$ pellet might exceed this industry accepted limit at SCWR conditions. Therefore, alternative fuels with higher thermal conductivities need to be investigated for SCWR use. Uranium carbide (UC), uranium nitride (UN), and uranium dicarbide $(UC2)$ are excellent fuel choices as they all have higher thermal conductivities compared with conventional nuclear fuels such as $UO2$, mixed oxides (MOX), and thoria $(ThO2)$. Inconel-600 has been selected as the sheath material due its high corrosion resistance and high yield strength in aggressive supercritical water (SCW) at high-temperatures. This paper presents the thermalhydraulics calculations of a generic PT-type SCWR fuel channel with a 43-element Inconel-600 bundle with UC and $UC2$ fuels. The bulk-fluid, sheath and fuel centerline temperature profiles, together with a heat transfer coefficient profile, were calculated for a generic PT-type SCWR fuel-bundle string. Fuel bundles with UC and $UC2$ fuels with various axial heat flux profiles (AHFPs) are acceptable since they do not exceed the sheath-temperature design limit of $850°C$, and the industry accepted limit for the fuel centerline temperature of $1850°C$. The most desirable case in terms of the lowest fuel centerline temperature is the UC fuel with the upstream-skewed cosine AHFP. In this case, the fuel centerline temperature does not exceed even the sheath-temperature design limit of $850°C$.

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## Figures

Figure 1

Temperature and HTC profiles of UO2 along the heated length of the fuel channel (centerline fuel temperature based on the average thermal conductivity of UO2) (5)

Figure 2

Nonuniform AHFPs (7)

Figure 3

Comparison of the thermal conductivities of the nuclear fuels (data for UC2 and ThO2 are taken from Refs. 8-9, and Ref. 6, respectively

Figure 4

Bulk-fluid temperature and thermophysical properties profiles along the heated-bundle length with downstream-skewed AHFP

Figure 5

Thermophysical properties profiles at downstream-skewed cosine AHFP: (a) “regular” and average specific heats, (b) regular and average Prandtl numbers, and (c) density ratio

Figure 6

Temperature and HTC profiles for UO2 fuel: (a) at cosine AHFP and (b) at downstream-skewed AHFP

Figure 7

Temperature and HTC profiles along the heated length of the fuel channel at uniform AHFP (6): (a) UC fuel and (b) UC2 fuel

Figure 8

Temperature and HTC profiles along the heated length of the fuel channel for UC2 fuel (14): (a) at upstream-skewed cosine AHFP, (b) at cosine AHFP, and (c) at downstream-skewed cosine AHFP

Figure 9

Temperature and HTC profiles along the heated length of the fuel channel for UC fuel (14): (a) at upstream-skewed cosine AHFP, (b) at cosine AHFP, and (c) at downstream-skewed cosine AHFP

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