In the past, stainless steel was utilized as cladding in many PWRs （pressurized water reactors）, and its performance under irradiation was excellent. However, stainless steel was replaced by zirconium-based alloy as...In the past, stainless steel was utilized as cladding in many PWRs （pressurized water reactors）, and its performance under irradiation was excellent. However, stainless steel was replaced by zirconium-based alloy as cladding material mainly due to its lower neutron absorption cross section. Now, stainless steel cladding appears as a possible solution for safety problems related to hydrogen production and explosion as occurred in Fukushima Daiichi accident. The aim of this paper is to discuss the steady-state irradiation performance using stainless steel as cladding. The results show that stainless steel rods display higher fuel temperatures and wider pellet-cladding gaps than Zircaloy rods and no gap closure. The thermal performance of the two rods is very similar and the neutron absorption penalty due to stainless steel use could be compensating by combining small increase in U-235 enrichment and pitch size changes.展开更多
The subject of this study is the oxidation of fuel rod cladding made of material Zr1Nb(0.1% O) in steam at temperatures in the range of 660℃ to 1200℃ with a surface in the initial state (after manufacturing - grindi...The subject of this study is the oxidation of fuel rod cladding made of material Zr1Nb(0.1% O) in steam at temperatures in the range of 660℃ to 1200℃ with a surface in the initial state (after manufacturing - grinding) and after additional chemical etching. The changes in the microstructure of tubes due to the interaction with steam were investigated. A comparison was made between the oxidation rate of this material (weight gain) and the data on the oxidation of other alloys for nuclear power plants. The oxidation rate of Zr1Nb(0.1% O) is close to the oxidation rate of other zirconium alloys. It is shown that after chemical treatment of the surface of the samples there is a more even growth of oxide films, and they have a smaller thickness for the same time of exposure than after mechanical grinding. Surface treatment before oxidation also affects the change of microstructure of samples when heated to high temperatures.展开更多
Nuclear fuel performance modeling and simulation are critical tasks for nuclear fuel design optimization and safety analysis under normal and transient conditions.Fuel performance is a complicated phenomenon that invo...Nuclear fuel performance modeling and simulation are critical tasks for nuclear fuel design optimization and safety analysis under normal and transient conditions.Fuel performance is a complicated phenomenon that involves thermal,mechanical,and irradiation mechanisms and requires special multiphysics modules.In this study,a fuel performance model was developed using the COMSOL Multiphysics platform.The modeling was performed for a 2D axis-symmetric geometry of a UO2fuel pellet in the E110 clad for VVER-1200 fuel.The modeling considers all relevant phenomena,including heat generation and conduction,gap heat transfer,elastic strain,mechanical contact,thermal expansion,grain growth,densification,fission gas generation and release,fission product swelling,gap/plenum pressure,and cladding thermal and irradiation creep.The model was validated using a code-to-code evaluation of the fuel pellet centerline and surface temperatures in the case of constant power,in addition to validation of fission gas release(FGR)predictions.This prediction proved that the model could perform according to previously published VVER nuclear fuel performance parameters.A sensitivity study was also conducted to assess the effects of uncertainty on some of the model parameters.The model was then used to predict the VVER-1200 fuel performance parameters as a function of burnup,including the temperature profiles,gap width,fission gas release,and plenum pressure.A compilation of related material and thermomechanical models was conducted and included in the modeling to allow the user to investigate different material/performance models.Although the model was developed for normal operating conditions,it can be modified to include off-normal operating conditions.展开更多
文摘In the past, stainless steel was utilized as cladding in many PWRs （pressurized water reactors）, and its performance under irradiation was excellent. However, stainless steel was replaced by zirconium-based alloy as cladding material mainly due to its lower neutron absorption cross section. Now, stainless steel cladding appears as a possible solution for safety problems related to hydrogen production and explosion as occurred in Fukushima Daiichi accident. The aim of this paper is to discuss the steady-state irradiation performance using stainless steel as cladding. The results show that stainless steel rods display higher fuel temperatures and wider pellet-cladding gaps than Zircaloy rods and no gap closure. The thermal performance of the two rods is very similar and the neutron absorption penalty due to stainless steel use could be compensating by combining small increase in U-235 enrichment and pitch size changes.
文摘The subject of this study is the oxidation of fuel rod cladding made of material Zr1Nb(0.1% O) in steam at temperatures in the range of 660℃ to 1200℃ with a surface in the initial state (after manufacturing - grinding) and after additional chemical etching. The changes in the microstructure of tubes due to the interaction with steam were investigated. A comparison was made between the oxidation rate of this material (weight gain) and the data on the oxidation of other alloys for nuclear power plants. The oxidation rate of Zr1Nb(0.1% O) is close to the oxidation rate of other zirconium alloys. It is shown that after chemical treatment of the surface of the samples there is a more even growth of oxide films, and they have a smaller thickness for the same time of exposure than after mechanical grinding. Surface treatment before oxidation also affects the change of microstructure of samples when heated to high temperatures.
基金The Science,Technology&Innovation Funding Authority(STDF)in cooperation with The Egyptian Knowledge Bank(EKB).
文摘Nuclear fuel performance modeling and simulation are critical tasks for nuclear fuel design optimization and safety analysis under normal and transient conditions.Fuel performance is a complicated phenomenon that involves thermal,mechanical,and irradiation mechanisms and requires special multiphysics modules.In this study,a fuel performance model was developed using the COMSOL Multiphysics platform.The modeling was performed for a 2D axis-symmetric geometry of a UO2fuel pellet in the E110 clad for VVER-1200 fuel.The modeling considers all relevant phenomena,including heat generation and conduction,gap heat transfer,elastic strain,mechanical contact,thermal expansion,grain growth,densification,fission gas generation and release,fission product swelling,gap/plenum pressure,and cladding thermal and irradiation creep.The model was validated using a code-to-code evaluation of the fuel pellet centerline and surface temperatures in the case of constant power,in addition to validation of fission gas release(FGR)predictions.This prediction proved that the model could perform according to previously published VVER nuclear fuel performance parameters.A sensitivity study was also conducted to assess the effects of uncertainty on some of the model parameters.The model was then used to predict the VVER-1200 fuel performance parameters as a function of burnup,including the temperature profiles,gap width,fission gas release,and plenum pressure.A compilation of related material and thermomechanical models was conducted and included in the modeling to allow the user to investigate different material/performance models.Although the model was developed for normal operating conditions,it can be modified to include off-normal operating conditions.
