Thermal Aspects of Uranium Carbide and Uranium ...

Thermal Aspects of Uranium Carbide and Uranium Dicarbide Fuels in Supercritical Water-Cooled Nuclear Reactors Lisa Grande e-mail: [email protected]

Bryan Villamere e-mail: [email protected]

Leyland Allison Sally Mikhael Adrianexy Rodriguez-Prado Igor Pioro e-mail: [email protected] Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, Ontario, L1H 7K4, Canada

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 共UO2兲. The sheathtemperature 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 共UC2兲 are excellent fuel choices as they all have higher thermal conductivities compared with conventional nuclear fuels such as UO2, mixed oxides (MOX), and thoria 共ThO2兲. Inconel-600 has been selected as the sheath material due its high corrosion resistance and high yield strength in aggressive supercritical water (SCW) at hightemperatures. This paper presents the thermalhydraulics calculations of a generic PTtype 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. 关DOI: 10.1115/1.4001299兴 Keywords: SCWR, uranium carbide, uranium dicarbide, thermalhydraulics, AHFP

1

Introduction

In the late 1950s and 1960s, USA and Russia began studying concepts of supercritical water-cooled nuclear reactors 共SCWRs兲 with the objective of increasing the thermal efficiency of nuclear power plants 共NPPs兲. After a nearly 30-year break, this idea became attractive again. Today, six Generation IV nuclear-reactor concepts, including the SCWR option, are being developed worldwide 关1兴. A SCWR concept studied in this paper is a pressure tube 共PT兲 or pressure channel 共PCh兲 type reactor. This reactor type utilizes pressure tubes or pressurized channels instead of a large pressure vessel. The main advantages of supercritical water 共SCW兲 NPPs Contributed by the Nuclear Division of ASME for publicatin in the JOURNAL OF ENGINEERING FOR GAS TURBINES AND POWER. Manuscript received January 18, 2010; final manuscript received January 21, 2010; published online October 25, 2010. Editor: Dilip R. Ballal.

关1–3兴 are as follows: 共1兲 improved thermal efficiency by 10–15% compared with current NPPs; 共2兲 decreased operational and capital costs, and reduced overall electrical energy cost; and 共3兲 a possibility for cogeneration of hydrogen. Supercritical water-cooled reactor technology is currently in its early design phase. A demonstration unit has yet to be designed and constructed. Fuel materials and configurations suited at supercritical conditions are currently being studied. This paper describes the thermal-design options of fuel bundles with respect to the maximum fuel centerline temperature 共CL兲 to be restricted to 1850° C, and the maximum sheath temperature 共SH兲 to be restricted to 850° C 关4兴. The model used in the current thermal-design analysis is a generic PT-type SCWR with 300 fuel channels and 1200-MWel power. A heated-channel length of 5.772 m is assumed. The anticipated fuel string consists of 12 bundles. Calculations consider the fuel-bundle string length to be equal to the heated-channel length, i.e., bundle end-plates and end-caps are not considered.

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Fig. 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 Fig. 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‡

Pressure drop along the channel was not accounted for, and pressure was assumed to be a constant of 25 MPa. The contact resistance between the fuel pellet and sheath was considered to be negligible. Steady-state conditions with uniform and several cosine axial heat flux profiles 共AHFPs兲 were applied. A coolant mass-flow rate per channel was assumed to be a constant of 4.4 kg/s and the produced power per channel to be 8.5 MWth.

2

Background

A previous study 关5兴 was performed to analyze the different design features in SCW PT-type nuclear reactors. This study was performed using a generic 43-element fuel bundle with UO2 fuel. However, this study considered only preliminary steady-state heat transfer calculations, with a uniform AHFP and an average fuel thermal conductivity. This study has shown that with the UO2 fuel, a fuel centerline temperature might exceed the industry accepted limit of 1850° C 共see Fig. 1兲. The same results were obtained by Allison et al. 关6兴. Therefore, the present paper is dedicated to more representative nuclear-reactor AHFPs, such as cosine, and upstream-skewed and downstream-skewed cosine profiles 共for details, see Fig. 2.兲, with two alternative fuels uranium carbide 共UC兲 and uranium dicarbide

共UC2兲, because these fuels have significantly higher thermal conductivities compared with those of UO2, mixed oxides 共MOX兲, and thoria 共ThO2兲 fuels.

3

Nuclear Fuels

The fuels compared in the present paper are UO2, UC, and UC2. The main objective is to achieve a fuel composition with a lower fuel centerline temperature suited for SCWR use. Uranium dioxide is the commonly used fuel in current reactors. However, it has a very low thermal conductivity that decreases as the temperature increases 共for details, see Fig. 3.兲. Therefore, new alternative nuclear fuels with higher thermal conductivities have to be considered. As shown in Fig. 3, thermal conductivities of UC, uranium nitride 共UN兲, and UC2 fuels are much higher than that of conventional nuclear fuels such as UO2, MOX, and ThO2, and their thermal conductivities increase with increasing temperature. A fuel with a rising trend in thermal conductivity would increase the heat transfer through a pellet and decrease the fuel centerline temperature. This rising trend in the thermal conductivity would be a key safety factor for SCWRs. Table 1 lists important thermophysical properties of nuclear fuels. In general, there are many parameters such as the density, porosity, method of manufacturing, etc., which might affect the thermal conductivity of any potential fuel 关8兴. Therefore, only generic thermal conductivities of nuclear fuels were used in the calculations to follow. 3.1 Property Profiles. Figure 4 shows thermophysical property profiles 共calculations based on the National Institute of Standards and Technology 共NIST兲 software 关11兴兲 of a light-water coolant along the heated-channel length for the downstream-skewed cosine AHFP. All thermophysical properties undergo significant and drastic changes within the pseudocritical 共PC兲 region. This statement applies also to all presented AHFPs. The only difference is that the pseudocritical-point location along the bundle string heated length will depend on the particular AHFP 共see also Refs. 关6,12兴兲. The average specific heat, average Prandtl number, and density ratio 共see Fig. 5兲 were used in the correlation of Bishop et al. 关13兴.

4

Fig. 2 Nonuniform AHFPs †7‡

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Methodology and Calculations

The thermophysical properties of the coolant at the sheath temperature and the thermal conductivities of the sheath and fuel were calculated using an iterative method. In general, coolant properties were estimated based on a bulk-fluid temperature, i.e., an average Transactions of the ASME


Table 1 Thermophysical properties of ceramic nuclear fuels at 0.1 MPa and 25° C †8,10‡ „only ThO2… Fuel Property Molar mass Theoretical density Melting temperature Boiling temperature Heat of fusion Specific heat Thermal conductivity Coefficient of linear expansion a

a

Units

UO2

MOX

ThO2

UN

UC

UC2

kg/kmol kg/ m3 °C °C kJ/kg kJ/kg K W/m K

270.3 10,960 2850⫾ 30 3542 259⫾ 15 0.235 8.68

271.2 11,074 2750 3538 285.3 0.240 7.82b

264 10,000 3227⫾ 150 ⬎4227 0.235 9.7

252 14,300 2850⫾ 30 0.190 13.0

250 13,630 2365⫾ 165 4418 195.6 0.200 25.3

262 11,700 2800⫾ 30 0.162 13

1/K

9.75⫻ 10−6

-

8.9⫻ 10−6

7.52⫻ 10−6

10.1⫻ 10−6

-

MOX 共U0.8Pu0.2兲O2, where 0.8 and 0.2 are the molar parts of UO2 and PuO2, respectively. At 95% density.

b

coolant temperature in a cross section. All calculations were performed along the heated-bundle length with a 1-mm increment. The bulk-fluid temperature was calculated through the heatbalance method. With the bulk-fluid temperature known and sheath temperature assumed initially, all thermophysical properties can be determined at these temperatures, and using the correlation of Bishop et al. 共see Eq. 共1兲兲, the heat transfer coefficient 共HTC兲 can be calculated based on iterations. The correlation of Bishop et al. is suitable within a pressure range from 22.8 MPa to 27.6 MPa, bulk-fluid temperatures between 282° C and 527° C, and Heat Flux 共HF兲 between 0.31 MW/ m2 and 3.46 MW/ m2 关13兴 Nux = 0.0069 RexPrx0.66

冉冊 ␳o,sh ␳b

0.43

共1兲 x

The correlation of Bishop et al. is applicable for tubes, and the last term 共1 + 2.4D / x兲 represents the entrance effect in a bare tube. This term can be neglected when accounting for the various fuelbundle appendages 共endplates, spacers, etc.兲. It was assumed a perfect contact between the fuel pellet and sheath. Therefore, the temperature of the inner-sheath surface equals to the temperature of the outer surface of a fuel pellet. The

fuel centerline temperature was determined by small radial increments with variable thermal conductivity.

5

Results

5.1 UO2 Fuel Centerline Temperature. Uranium dioxide fuel centerline temperature surpasses the industry accepted limit of 1850° C at uniform 共see Fig. 1兲, cosine 共see Fig. 6共a兲兲, and downstream-skewed cosine AHFPs 共see Fig. 6共b兲兲. 5.2 UC and UC2 Fuel Centerline Temperatures. Figure 7 shows variations in the temperatures and HTC profiles along the heated-bundle length at uniform AHFP for UC and UC2 fuels. Figures 8 and 9 shows the temperatures and HTC profiles for the UC2 fuel and UC fuel at nonuniform AHFPs, respectively. An analysis of these graphs shows that all calculated cases with UC and UC2 have significantly lower fuel centerline temperatures compared with those of UO2 fuel at all uniform and nonuniform AHFPs due to their higher thermal conductivity. 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. The least

Fig. 4 Bulk-fluid temperature and thermophysical properties profiles along the heated-bundle length with downstream-skewed AHFP

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Fig. 6 Temperature and HTC profiles for UO2 fuel: „a… at cosine AHFP and „b… at downstream-skewed AHFP

With both UC and UC2 fuels there are factors beyond the scope of this paper, with regard to the porosity, density, and manufacturing process, which must be accounted for when determining the feasibility of these substances as nuclear fuels for SCWRs. In all the analyzed cases the Iconel-600 sheath temperature is below the design limit of 850° C.

6

Fig. 5 Thermophysical properties profiles at downstreamskewed cosine AHFP: „a… “regular” and average specific heats, „b… regular and average Prandtl numbers, and „c… density ratio

desirable case in terms of the fuel centerline temperature is UC2 with downstream-skewed cosine AFHP. This “worse” case is still acceptable, because the fuel centerline temperature is significantly less than the industry accepted temperature limit of 1850° C. 022901-4 / Vol. 133, FEBRUARY 2011

Conclusions

In general, UO2 nuclear fuel might not be a good choice for SCWRs, because at certain conditions, the fuel centerline temperature exceeds the industry accepted limit of 1850° C. Uranium nitride, UC, and UC2 nuclear fuels with significantly higher thermal conductivities might be considered as alternative fuels compared with conventional fuels such as UO2, MOX, and ThO2. These fuels can be used at any AHFPs: uniform, cosine, upstream-skewed or downstream-skewed cosine. However, further material investigation would be required into the properties of UN, UC, and UC2 fuels. The uranium carbide nuclear fuel, with the highest thermal conductivity values compared with that of other fuels 共UO2, MOX, ThO2, UN, and UC2兲 will have the largest safety margin for the fuel centerline temperature. Transactions of the ASME


Fig. 7 Temperature and HTC profiles along the heated length of the fuel channel at uniform AHFP †6‡: „a… UC fuel and „b… UC2 fuel

Acknowledgment Financial supports from the NSERC Discovery Grant and NSERC/NRCan/AECL Generation-IV Energy Technologies Program are gratefully acknowledged.

Nomenclature Afl D cp ¯c p

⫽ ⫽ ⫽ ⫽

G h HTC k m P Q T x

⫽ ⫽ ⫽ ⫽ ⫽ ⫽ ⫽ ⫽ ⫽

flow area 共m2兲 diameter 共m兲 specific heat 共J/kg K兲 average specific heat, 共共ho,sh − hb兲 / 共To,sh − Tb兲兲共J/kg K兲 ˙ / Aflow兲共kg/ m2 s兲 mass flux, 共m enthalpy 共J/kg兲 heat transfer coefficient 共W / m2 K兲 thermal conductivity 共W/m K兲 mass-flow rate 共kg/s兲 pressure 共Pa兲 heat transfer rate 共W兲 temperature 共°C兲 axial coordinate 共m兲


Fig. 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 downstreamskewed cosine AHFP

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Greek Symbols ␮ ⫽ dynamic viscosity 共Pa s兲 ␳ ⫽ density 共kg/ m3兲 Nondimensional Numbers Nu ⫽ Nusselt number, 共HTCDhy / k兲 Pr ⫽ Prandtl number, 共c p␮ / k兲 Pr ⫽ average Prandtl number, 共c ¯ p␮ / k兲 Re ⫽ Reynolds number, 共GDhy / ␮兲 Subscripts b ch el hy max o pc sh th w

⫽ ⫽ ⫽ ⫽ ⫽ ⫽ ⫽ ⫽ ⫽ ⫽

Abbreviations AHFP ⫽ CL ⫽ HF ⫽ HTC ⫽ MOX ⫽ NIST ⫽ NPP ⫽ OD ⫽ PC ⫽ PCh ⫽ PT ⫽ SCW ⫽ SCWR ⫽

bulk channel electrical hydraulic-equivalent maximum outer pseudocritical sheath thermal wall axial heat flux profile fuel centerline temperature heat flux heat transfer coefficient mixed oxide National Institute of Standards and Technology nuclear power plant outer diameter pseudocritical pressure channel pressure tube supercritical water supercritical water-cooled reactor

References

Fig. 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 downstreamskewed cosine AHFP

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Transactions of the ASME


关12兴 Grande, L., Villamere, B., Rodriguez-Prado, A., Mikhael, S., Allison, L., and Pioro, I., 2009, “Thermal Aspects of Using Thoria Fuel in Supercritical WaterCooled Nuclear Reactors,” Proceedings of the ICONE-17, Brussels, Belgium, Jul. 12–16, Paper No. 75969. 关13兴 Bishop, A., Sandberg, R., and Tong, L., 1964, “Forced Convection Heat Transfer to Water at Near-Rcritical Temperatures and Super-Critical Pressures,”


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