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.

1.
Pioro
,
I.
, and
Duffey
,
R.
, 2007,
Heat Transfer and Hydraulic Resistance at Supercritical Pressures in Power Engineering Applications
,
ASME
,
New York
.
2.
Naidin
,
M.
,
Pioro
,
I.
,
Zirn
,
U.
,
Mokry
,
S.
, and
Naterer
,
G.
, 2009a, “
Supercritical Water-Cooled NPPs With Co-Generation of Hydrogen: General Layout and Thermodynamic-Cycles Options
,”
Proceedings of the Fourth International Symposium on Supercritical Water-Cooled Reactors
, Heidelberg, Germany, Mar. 8–11, Paper No. 78.
3.
Naidin
,
M.
,
Mokry
,
S.
,
Baig
,
F.
,
Gospodinov
,
Y.
,
Zirn
,
U.
,
Pioro
,
I.
, and
Naterer
,
G.
, 2009, “
Thermal-Design Options for Pressure-Channel SCWRs With Co-Generation of Hydrogen
,”
ASME J. Eng. Gas Turbines Power
0742-4795,
131
, p.
012901
.
4.
Chow
,
C. K.
, and
Khartabil
,
H. F.
, 2007, “
Conceptual Fuel Channel Designs for CANDU-SCWR
,”
Nuclear Engineering and Technology: An International Journal of the Korean Nuclear Society
,
40
(
2
), pp.
77
84
.
5.
Pioro
,
I. L.
,
Khan
,
M.
,
Hopps
,
V.
,
Jacobs
,
Ch.
,
Patkunam
,
R.
,
Gopaul
,
S.
, and
Bakan
,
K.
, 2008, “
SCW Pressure Channel Nuclear Reactor, Some Design Features
,”
JSME J. of Power and Energy Systems
,
2
(
2
), pp.
874
888
.
6.
Allison
,
L.
,
Grande
,
L.
,
Villamere
,
B.
,
Mikhael
,
S.
,
Rodriguez-Prado
,
A.
, and
Pioro
,
I.
, 2009, “
Thermal Design Option Using Uranium Carbide and Uranium Dicarbide in SCWR Uniformally Heated Fuel Channel
,”
Proceedings of the ICONE-17
, Brussels, Belgium, Jul. 12–16, Paper No. 75975.
7.
Leung
,
L. K.
, 2008, “
Effect of CANDU Bundle-Geometry Variation on Dryout Power
,”
Proceedings of the ICONE-16
, Orlando, FL, Paper No. 48827.
8.
Kirillov
,
P.
,
Terent'eva
,
M.
, and
Deniskina
,
N.
, 2007,
Thermophsical Properties of Nuclear Engineering Materials (in Russian)
,
2nd ed.
,
IzdAT Publishing House
,
Moscow, Russia
.
9.
Chirkin
,
V.
, 1968,
Thermophysical Properties of Materials for Nuclear Engineering (in Russian)
,
Atomizdat Publishing House
,
Moscow, Russia
.
10.
Jain
,
D.
,
Pillai
,
C.
,
Rao
,
B. K.
, and
Sahoo
,
K.
, 2006, “
Thermal Diffusivity and Thermal Conductivity of Thoria-Lanthana Solid Solutions Up to 10 mol % LaO(1.5)
,”
J. Nucl. Mater.
0022-3115,
353
, pp.
35
41
.
11.
National Institute of Standards and Technology
, 2007,
NIST Reference Fluid Thermodynamic and Transport Properties-REFPROP, NIST Standard Reference Database 23 Ver. 8.0
,
Department of Commerce
,
Boulder, CO
.
12.
Grande
,
L.
,
Villamere
,
B.
,
Rodriguez-Prado
,
A.
,
Mikhael
,
S.
,
Allison
,
L.
, and
Pioro
,
I.
, 2009, “
Thermal Aspects of Using Thoria Fuel in Supercritical Water-Cooled Nuclear Reactors
,”
Proceedings of the ICONE-17
, Brussels, Belgium, Jul. 12–16, Paper No. 75969.
13.
Bishop
,
A.
,
Sandberg
,
R.
, and
Tong
,
L.
, 1964, “
Forced Convection Heat Transfer to Water at Near-Rcritical Temperatures and Super-Critical Pressures
,” Westinghouse Electric Corporation, Atomic Power Division, Pittsburgh, PA, Report No. WCAP-2056.
14.
Villamere
,
B.
,
Grande
,
L.
,
Rodriguez-Prado
,
A.
,
Mikhael
,
S.
,
Allison
,
L.
, and
Pioro
,
I.
, 2009, “
Thermal Aspects for Uranium Carbide and Uranium Dicarbide Fuels in Supercritical Water-Cooled Nuclear Reactors
,”
Proceedings of the ICONE-17
, Brussels, Belgium, Jul. 12–16, Paper No. 75990.
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