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Title: An Innovative High Thermal Conductivity Fuel Design

Abstract

Uranium dioxide (UO2) is the most common fuel material in commercial nuclear power reactors. UO2 has the advantages of a high melting point, good high-temperature stability, good chemical compatibility with cladding and coolant, and resistance to radiation. The main disadvantage of UO2 is its low thermal conductivity. During a reactor’s operation, because the thermal conductivity of UO2 is very low, for example, about 2.8 W/m-K at 1000 oC [1], there is a large temperature gradient in the UO2 fuel pellet, causing a very high centerline temperature, and introducing thermal stresses, which lead to extensive fuel pellet cracking. These cracks will add to the release of fission product gases after high burnup. The high fuel operating temperature also increases the rate of fission gas release and the fuel pellet swelling caused by fission gases bubbles. The amount of fission gas release and fuel swelling limits the life time of UO2 fuel in reactor. In addition, the high centerline temperature and large temperature gradient in the fuel pellet, leading to a large amount of stored heat, increase the Zircaloy cladding temperature in a lost of coolant accident (LOCA). The rate of Zircaloy-water reaction becomes significant at the temperature above 1200 oC [2].more » The ZrO2 layer generated on the surface of the Zircaloy cladding will affect the heat conduction, and will cause a Zircaloy cladding rupture. The objective of this research is to increase the thermal conductivity of UO2, while not affecting the neutronic property of UO2 significantly. The concept to accomplish this goal is to incorporate another material with high thermal conductivity into the UO2 pellet. Silicon carbide (SiC) is a good candidate, because the thermal conductivity of single crystal SiC is 60 times higher than that of UO2 at room temperature and 30 times higher at 800 oC [3]. Silicon carbide also has the properties of low thermal neutron absorption cross section, high melting point, good chemical stability and good irradiation stability. Silicon carbide is expected to form a conductive lattice in UO2 for heat to flow out of the fuel pellet, and the thermal conductivity of SiC is anticipated to control the thermal conductivity of the fuel pellet. In this research, the effect of the SiC additive on the neutronic properties of a UO2 pellet was simulated by CASMO-3, a multi-group two-dimensional transport theory code. Three methods were studied to incorporate SiC into UO2. Firstly, silicon carbide whiskers were mixed with UO2 particles and then hot press sintered to achieve dense pellets. Secondly, a polymer precursor, allylhydridopolycarbosilane (AHPCS), was used to generate a SiC coating on UO2 particles prior to the hot press sintering process. Thirdly, chemical vapor deposition (CVD) process was used to coat UO2 particles with a SiC layer prior to the sintering process. To avoid a reaction that occurs between UO2 and SiC at 1377 oC [4], a two stages low temperature sintering method was used to sinter the mixture of SiC whiskers and UO2 particles or the SiC coated UO2 particles at 1200 oC. The sintered pellets were characterized by X-ray diffraction (XRD) and Scanning Electron Microscope (SEM), and the thermal conductivity of the sintered pellets was to be measured by laser flash method at Idaho National Laboratory. The centerline temperatures of the sintered pellets at the reactor operating condition were calculated based on the measured thermal conductivity.« less

Authors:
;
Publication Date:
Research Org.:
University of Florida
Sponsoring Org.:
USDOE - Office of Nuclear Energy, Science and Technology (NE)
OSTI Identifier:
917989
Report Number(s):
DOE F 241 3
UF # 00050571; TRN: US0807336
DOE Contract Number:  
FG07-04ID14598
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; ELECTRON MICROSCOPES; FISSION; FISSION PRODUCTS; FUEL PELLETS; NUCLEAR POWER; PELLETS; SILICON CARBIDES; SINTERING; STABILITY; THERMAL CONDUCTIVITY; THERMAL NEUTRONS; THERMAL STRESSES; URANIUM DIOXIDE; X-RAY DIFFRACTION; ZIRCALOY; commercial nuclear power fuel Uranium dioxide silicon carbide thermal conductivity

Citation Formats

PI: James S. Tulenko, and Co-PI: Ronald H. Baney,. An Innovative High Thermal Conductivity Fuel Design. United States: N. p., 2007. Web. doi:10.2172/917989.
PI: James S. Tulenko, & Co-PI: Ronald H. Baney,. An Innovative High Thermal Conductivity Fuel Design. United States. doi:10.2172/917989.
PI: James S. Tulenko, and Co-PI: Ronald H. Baney,. Sun . "An Innovative High Thermal Conductivity Fuel Design". United States. doi:10.2172/917989. https://www.osti.gov/servlets/purl/917989.
@article{osti_917989,
title = {An Innovative High Thermal Conductivity Fuel Design},
author = {PI: James S. Tulenko and Co-PI: Ronald H. Baney,},
abstractNote = {Uranium dioxide (UO2) is the most common fuel material in commercial nuclear power reactors. UO2 has the advantages of a high melting point, good high-temperature stability, good chemical compatibility with cladding and coolant, and resistance to radiation. The main disadvantage of UO2 is its low thermal conductivity. During a reactor’s operation, because the thermal conductivity of UO2 is very low, for example, about 2.8 W/m-K at 1000 oC [1], there is a large temperature gradient in the UO2 fuel pellet, causing a very high centerline temperature, and introducing thermal stresses, which lead to extensive fuel pellet cracking. These cracks will add to the release of fission product gases after high burnup. The high fuel operating temperature also increases the rate of fission gas release and the fuel pellet swelling caused by fission gases bubbles. The amount of fission gas release and fuel swelling limits the life time of UO2 fuel in reactor. In addition, the high centerline temperature and large temperature gradient in the fuel pellet, leading to a large amount of stored heat, increase the Zircaloy cladding temperature in a lost of coolant accident (LOCA). The rate of Zircaloy-water reaction becomes significant at the temperature above 1200 oC [2]. The ZrO2 layer generated on the surface of the Zircaloy cladding will affect the heat conduction, and will cause a Zircaloy cladding rupture. The objective of this research is to increase the thermal conductivity of UO2, while not affecting the neutronic property of UO2 significantly. The concept to accomplish this goal is to incorporate another material with high thermal conductivity into the UO2 pellet. Silicon carbide (SiC) is a good candidate, because the thermal conductivity of single crystal SiC is 60 times higher than that of UO2 at room temperature and 30 times higher at 800 oC [3]. Silicon carbide also has the properties of low thermal neutron absorption cross section, high melting point, good chemical stability and good irradiation stability. Silicon carbide is expected to form a conductive lattice in UO2 for heat to flow out of the fuel pellet, and the thermal conductivity of SiC is anticipated to control the thermal conductivity of the fuel pellet. In this research, the effect of the SiC additive on the neutronic properties of a UO2 pellet was simulated by CASMO-3, a multi-group two-dimensional transport theory code. Three methods were studied to incorporate SiC into UO2. Firstly, silicon carbide whiskers were mixed with UO2 particles and then hot press sintered to achieve dense pellets. Secondly, a polymer precursor, allylhydridopolycarbosilane (AHPCS), was used to generate a SiC coating on UO2 particles prior to the hot press sintering process. Thirdly, chemical vapor deposition (CVD) process was used to coat UO2 particles with a SiC layer prior to the sintering process. To avoid a reaction that occurs between UO2 and SiC at 1377 oC [4], a two stages low temperature sintering method was used to sinter the mixture of SiC whiskers and UO2 particles or the SiC coated UO2 particles at 1200 oC. The sintered pellets were characterized by X-ray diffraction (XRD) and Scanning Electron Microscope (SEM), and the thermal conductivity of the sintered pellets was to be measured by laser flash method at Idaho National Laboratory. The centerline temperatures of the sintered pellets at the reactor operating condition were calculated based on the measured thermal conductivity.},
doi = {10.2172/917989},
journal = {},
number = ,
volume = ,
place = {United States},
year = {2007},
month = {10}
}