Accident-tolerant fuel(ATF)has attracted considerable research attention since the 2011 Fukushima nuclear disaster.To improve the accident tolerance of the fuel-cladding systems in the current light-water reactors,it ...Accident-tolerant fuel(ATF)has attracted considerable research attention since the 2011 Fukushima nuclear disaster.To improve the accident tolerance of the fuel-cladding systems in the current light-water reactors,it is proposed to develop and deploy(1)an enhanced Zrbased alloy or coated zircaloy for the fuel cladding,(2)alternative cladding materials with better accident tolerance,and(3)alternative fuels with enhanced accident tolerance and/or a higher U density.This review presents the features of the current UO2-zircaloy system.Different techniques and characters to develop coating materials and enhanced Zr-based alloys are summarized.The features of several selected alternative fuels and cladding materials are reviewed and discussed.The neutronic evaluations of alternative fuel-cladding systems are analyzed.It is expected that one or more types of ATF-cladding systems discussed in the present review will be implemented in commercial reactors.展开更多
In resonance with the Fukushima Daiichi Nuclear Power Plant accident lesson, a novel fuel design to enhance safety regarding severe accident scenarios has become increasingly appreciated in the nuclear power industry....In resonance with the Fukushima Daiichi Nuclear Power Plant accident lesson, a novel fuel design to enhance safety regarding severe accident scenarios has become increasingly appreciated in the nuclear power industry. This research focuses on analysis of the neutronic properties of a silicon carbide(SiC) cladding fuel assembly, which provides a greater safety margin as a type of accident-tolerant fuel for pressurized water reactors. The general physical performance of SiC cladding is explored to ascertain its neutronic performance. The neutron spectrum, accumulation of ^(239)Pu, physical characteristics,temperature reactivity coefficient, and power distribution are analyzed. Furthermore, the influences of a burnable poison rod and enrichment are explored. SiC cladding assemblies show a softer neutron spectrum and flatter power distribution than conventional Zr alloy cladding fuel assemblies. Lower enrichment fuel is required when SiC cladding is adopted. However, the positive reactivity coefficient associated with the SiC material remains to be offset. The results reveal that SiC cladding assemblies show broad agreement with the neutronic performance of conventional Zr alloy cladding fuel. In the meantime, its unique physical characteristics can lead to improved safety and economy.展开更多
The temperature-dependent effective thermal conductivity of UN-X-UO_(2)(X=Mo,W)nuclear fuel composite was estimated.Following the experimental design,the thermal conductivity was calculated using Finite Element Modeli...The temperature-dependent effective thermal conductivity of UN-X-UO_(2)(X=Mo,W)nuclear fuel composite was estimated.Following the experimental design,the thermal conductivity was calculated using Finite Element Modeling(FEM),and compared with analytical models for 10%,30%,50%,and 70%(in mass)uncoated/coated UN microspheres in a UO2 matrix.The FEM results show an increase in the fuel thermal conductivity as the mass fraction of the UN microspheres increases from 1.2 to 4.6 times the UO2 reference at 2,000 K.The results from analytical models agree with the thermal conductivity estimated by FEM.The results also show that Mo and W coatings have similar thermal behaviors,and the coating thickness influences the thermal conductivity of the composite.At higher weight fractions,the thermal conductivity of the fuel composite at room temperature is substantially influenced by the high thermal conductivity coatings approaching that of UN.Thereafter,the thermal conductivity from FEM was used in the fuel thermal performance evaluation during LWR normal operation to calculate the maximum centerline temperature.The results show a significant decrease in the fuel maximum centerline temperature ranging from−94 K for 10% UN to−414 K for 70%(in mass)UN compared to UO2 under the same operating conditions.展开更多
基金supported by the National Key Research and Development Program of China(No.2018YFB1900405)the National Natural Science Foundation of China(No.11775316)+3 种基金the Tip-top Scientific and Technical Innovative Youth Talents of Guangdong Special Support Program(No.2016TQ03N575)the Fundamental Research Funds for the Central Universities(No.19lgpy299)the Science and Technology Planning Project of Guangdong Province,China(No.2019A050510022)the Nuclear Power Institute of China(No.HT-ATF-14-2018001)
文摘Accident-tolerant fuel(ATF)has attracted considerable research attention since the 2011 Fukushima nuclear disaster.To improve the accident tolerance of the fuel-cladding systems in the current light-water reactors,it is proposed to develop and deploy(1)an enhanced Zrbased alloy or coated zircaloy for the fuel cladding,(2)alternative cladding materials with better accident tolerance,and(3)alternative fuels with enhanced accident tolerance and/or a higher U density.This review presents the features of the current UO2-zircaloy system.Different techniques and characters to develop coating materials and enhanced Zr-based alloys are summarized.The features of several selected alternative fuels and cladding materials are reviewed and discussed.The neutronic evaluations of alternative fuel-cladding systems are analyzed.It is expected that one or more types of ATF-cladding systems discussed in the present review will be implemented in commercial reactors.
基金supported by the National Natural Science Foundation of China(No.11675057)the Fundamental Research Funds for the Central Universities(No.2017ZD100)
文摘In resonance with the Fukushima Daiichi Nuclear Power Plant accident lesson, a novel fuel design to enhance safety regarding severe accident scenarios has become increasingly appreciated in the nuclear power industry. This research focuses on analysis of the neutronic properties of a silicon carbide(SiC) cladding fuel assembly, which provides a greater safety margin as a type of accident-tolerant fuel for pressurized water reactors. The general physical performance of SiC cladding is explored to ascertain its neutronic performance. The neutron spectrum, accumulation of ^(239)Pu, physical characteristics,temperature reactivity coefficient, and power distribution are analyzed. Furthermore, the influences of a burnable poison rod and enrichment are explored. SiC cladding assemblies show a softer neutron spectrum and flatter power distribution than conventional Zr alloy cladding fuel assemblies. Lower enrichment fuel is required when SiC cladding is adopted. However, the positive reactivity coefficient associated with the SiC material remains to be offset. The results reveal that SiC cladding assemblies show broad agreement with the neutronic performance of conventional Zr alloy cladding fuel. In the meantime, its unique physical characteristics can lead to improved safety and economy.
基金This work was financially supported by the Swedish Science Council(Vetenskapsradet)under grant number 2019-04156by the Swedish Foundation for Strategic Research(SSF,Stiftelsen for Strategisk Forskning)under grant number ID17-0078,as well as in the SUNRISE center with financial support from SSF under Grant No.ARC19-0043.
文摘The temperature-dependent effective thermal conductivity of UN-X-UO_(2)(X=Mo,W)nuclear fuel composite was estimated.Following the experimental design,the thermal conductivity was calculated using Finite Element Modeling(FEM),and compared with analytical models for 10%,30%,50%,and 70%(in mass)uncoated/coated UN microspheres in a UO2 matrix.The FEM results show an increase in the fuel thermal conductivity as the mass fraction of the UN microspheres increases from 1.2 to 4.6 times the UO2 reference at 2,000 K.The results from analytical models agree with the thermal conductivity estimated by FEM.The results also show that Mo and W coatings have similar thermal behaviors,and the coating thickness influences the thermal conductivity of the composite.At higher weight fractions,the thermal conductivity of the fuel composite at room temperature is substantially influenced by the high thermal conductivity coatings approaching that of UN.Thereafter,the thermal conductivity from FEM was used in the fuel thermal performance evaluation during LWR normal operation to calculate the maximum centerline temperature.The results show a significant decrease in the fuel maximum centerline temperature ranging from−94 K for 10% UN to−414 K for 70%(in mass)UN compared to UO2 under the same operating conditions.
