This paper presents a corrosion assessment of copper spent nuclear fuel disposal canisters in crystalline rock,using hydrogeological modeling.A simplified approach is considered,to avoid complex and time-consuming com...This paper presents a corrosion assessment of copper spent nuclear fuel disposal canisters in crystalline rock,using hydrogeological modeling.A simplified approach is considered,to avoid complex and time-consuming computer simulations.This simplified case is presented as a base case,with changes in the hydrogeological parameters presented as variant cases.The results show that in Taiwan’s base case,decreasing the hydraulic conductivity of the rock or decreasing the hydraulic conductivity of dikes results in a shorter transport path for sulfide and an increase in corrosion depth.However,the estimated canister failure time is still over one million years in the variant cases.展开更多
<正>The disposal of spent nuclear fuel is a long-standing issue in nuclear technology.Mainly,UO_2 and metallic U arc used as a fuel in nuclear reactors.Spent nuclear fuel contains fission products and transurani...<正>The disposal of spent nuclear fuel is a long-standing issue in nuclear technology.Mainly,UO_2 and metallic U arc used as a fuel in nuclear reactors.Spent nuclear fuel contains fission products and transuranium elements,which would remain radioactive for 10~4 to 10~8 years.In this brief communication,essential concepts and engineering elements related to high-level nuclear waste disposal are described.Conceptual design models are described and discussed considering the long-time scale activity of spent nuclear fuel or high level waste.Notions of physical and chemical barriers to contain nuclear waste are highiightened.Concerns regarding integrity,self-irradiation induced decomposition and thermal effects of decay heat on the spent nuclear fuel are also discussed.The question of retrievability of spent nuclear fuel after disposal is considered.展开更多
To aid the United States Nuclear Regulatory Commission,Sandia National Laboratories (SNL) was contracted to investigate the seismic behavior of typical dry cask storage systems. Parametric evaluations characterized th...To aid the United States Nuclear Regulatory Commission,Sandia National Laboratories (SNL) was contracted to investigate the seismic behavior of typical dry cask storage systems. Parametric evaluations characterized the sensitivity of calculated cask response characteristics to input parameters. The parametric evaluation investigated two generic cask designs (cylindrical and rectangular),three different foundation types (soft soil,hard soil,and rock),and three different casks to pad coefficients of friction (0.2,0.55,0.8) for earthquakes with peak ground accelerations of 0.25g,0.6g,1.0g and 1.25g. A total of 1 165 analyses were completed,with regression analyses being performed on the resulting cask response data to determine relationships relating the mean (16 % and 84 % confidence intervals on the mean) to peak ground acceleration,peak ground velocity,and pseudo-spectral acceleration at 1 Hz and 5 % damping. In general,the cylindrical casks experienced significantly larger responses in comparison to the rectangular cask. The cylindrical cask experienced larger top of cask displacements,larger cask rotations (rocking),and more occurrences of cask toppling (the rectangular cask never toppled over). The cylindrical cask was also susceptible to rolling once rocking had been initiated,a behavior not observed in the rectangular cask. Cask response was not overly sensitive to foundation type,but was significantly dependent on the response spectrum employed.展开更多
Experience of operating reactor facilities (RF) with lead-bismuth coolant (LBC) has revealed that it is possible to perform safe refueling in short terms if the whole core is replaced and a kit of the special refuelin...Experience of operating reactor facilities (RF) with lead-bismuth coolant (LBC) has revealed that it is possible to perform safe refueling in short terms if the whole core is replaced and a kit of the special refueling equipment is used. However, comparing with RFs of nuclear submarines (NS), in which at the moment of performance of refueling the residual heat release is small, at RF SVBR-100 in a month after the reactor has been shut down, at the moment of performance of refueling the residual heat release is about 500 kW. Therefore, it is required to place the spent removable unit (SRU) with spent fuel subassemblies (SFSA) into the temporal storage tank (TST) filled with liquid LBC, in which the conditions for coolant natural circulation (NC) and heat removal via the tank vessel to the water cooling system are provided. After the residual heat release has been lowered to the level allowing transportation of the TST with SRU in the transporting-package container (TPC), it is proposed to consider a variant of TPCs transportation to the special site. On that site after the SRU has been reloaded into the long storage tank (LST) filled with quickly solidifying liquid lead, the TPCs can be stored during the necessary period. Thus, the controlled storage of LSTs is realized during several decades untill the time when SNF reprocessing and NFC closing are becoming economically expedient. On that storage, the four safety barriers are formed on the way of the release of radioactive products into the environment, namely: fuel matrix, fuel element cladding, solid lead and steel casing of the LST.展开更多
Dry storage containers must be secured and reliable during long-term storage,and the effect of decay heat released from the internal spent fuel on the cask has become an important research topic.In this paper,a 3D com...Dry storage containers must be secured and reliable during long-term storage,and the effect of decay heat released from the internal spent fuel on the cask has become an important research topic.In this paper,a 3D computational fluid dynamics model is presented,and the accuracy of the calculation is verified,with computational errors of less than 6.2%.The thermal stress of the dry storage cask was estimated by coupling it with a transient temperature field.The total power remained constant and adjusting the power ratio of the inner and outer zones had a small effect on the stress results,with a maximum equivalent stress of approximately 5.2 kPa,which occurred at the lower edge of the shell.In the case of tilt,the temperature gradient varied in a wavy distribution,and the wave crest moved from right to left.Altering the tilt angle affects the air distribution in the annular gap,leading to the shell temperature being transformed,with a maximum equivalent stress of 202 MPa at the bottom of the shell.However,the equivalent stress in both cases was less than the yield stress(205 MPa).展开更多
Reactor neutrino oscillation experiments, such as Daya Bay, Double Chooz and RENO are designed to determine the neutrino mixing angle θ13 with a sensitivity of 0.01--0.03 in sin^2 2θ13 at 90% confidence level, an im...Reactor neutrino oscillation experiments, such as Daya Bay, Double Chooz and RENO are designed to determine the neutrino mixing angle θ13 with a sensitivity of 0.01--0.03 in sin^2 2θ13 at 90% confidence level, an improvement over the current limit by more than one order of magnitude. The control of systematic uncertainties is critical to achieving the sin^22θ13 sensitivity goal of these experiments. Antineutrinos emitted from spent nuclear fuel (SNF) would distort the soft part of energy spectrum and may introduce a non-negligible systematic uncertainty. In this article, a detailed calculation of SNF neutrinos is performed taking account of the operation" of a typical reactor and the event rate in the detector is obtained. A further estimation shows that the event rate contribution of SNF neutrinos is less than 0.2% relative to the reactor neutrino signals. A global X2 analysis shows that this uncertainty will degrade the θ13 sensitivity at a negligible level.展开更多
The Daya Bay Reactor Antineutrino Experiment is designed to determine the as yet unknown neutrino mixing angle,θ13,by measuring the disappearance of electron antineutrinos from several nuclear reactor cores.The proje...The Daya Bay Reactor Antineutrino Experiment is designed to determine the as yet unknown neutrino mixing angle,θ13,by measuring the disappearance of electron antineutrinos from several nuclear reactor cores.The projected sensitivity in sin2(2θ13) of better than 0.01 at a 90% CL should be achieved after three years of data-taking.Antineutrinos emitted from spent nuclear fuel (SNF) distort the soft part of the energy spectrum.In this article,a calculation of the antineutrino spectra from the long-life isotopes in SNF is performed.A non-equilibrium generation of long half-life isotopes during the running time of the reactor is also analyzed.Finally,we show that the antineutrino event rate contribution from SNF,which has been stored in the SNF pool for several years,may be non-negligible.展开更多
A new physical simulation model was developed to simulate the casting process of the ductile iron heavy section spent-nuclear-fuel container.In this physical simulation model,a heating unit with DR24 Fe-Cr-Al heating ...A new physical simulation model was developed to simulate the casting process of the ductile iron heavy section spent-nuclear-fuel container.In this physical simulation model,a heating unit with DR24 Fe-Cr-Al heating wires was used to compensate the heat loss across the non-natural surfaces of the sample,and a precise and reliable casting temperature controlling/monitoring system was employed to ensure the thermal behavior of the simulated casting to be similar to the actual casting.Also,a mould system was designed,in which changeable mould materials can be used for both the outside and inside moulds for different applications.The casting test was carried out with the designed mould and the cooling curves of central and edge points at different isothermal planes of the casting were obtained.Results show that for most isothermal planes,the temperature control system can keep the temperature differences within 6 ℃ between the edge points and the corresponding center points,indicating that this new physical simulation model has high simulation accuracy,and the mould developed can be used for optimization of casting parameters of spent-nuclear-fuel container,such as composition of ductile iron,the pouring temperature,the selection of mould material and design of cooling system.In addition,to maintain the spheroidalization of the ductile iron,the force-chilling should be used for the current physical simulation to ensure the solidification of casting in less than 2 h.展开更多
High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CAND...High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (~22.5%), increase the energy output (~41%), decrease the quantity of spent fuels to be disposed (~2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.展开更多
For an analysis of the oxidation behavior of UO2 nuclear fuel pellet under a loss of water coolant accident in a spent nuclear fuel pool of an LWR, thermodynamic assessments of UO2 oxidation were carried out under var...For an analysis of the oxidation behavior of UO2 nuclear fuel pellet under a loss of water coolant accident in a spent nuclear fuel pool of an LWR, thermodynamic assessments of UO2 oxidation were carried out under various atmospheric conditions. In a steam atmosphere, it was assessed that UO2 would not be fully oxidized into U3O8 due to the relatively lower oxygen partial pressure, while UO2 will be fully oxidized into U3O8 in an air atmosphere. In an air and steam mixture atmosphere, the UO2 oxidation was dominantly affected by the air volumetric fraction, because of the relatively higher oxygen partial pressure of air. In addition, the effect of H2 volumetric fraction on the oxygen partial pressure under a mixture atmosphere was calculated, and it was revealed that UO2 pellet oxidation could be reduced above the critical value of H2 volumetric fraction.展开更多
文摘This paper presents a corrosion assessment of copper spent nuclear fuel disposal canisters in crystalline rock,using hydrogeological modeling.A simplified approach is considered,to avoid complex and time-consuming computer simulations.This simplified case is presented as a base case,with changes in the hydrogeological parameters presented as variant cases.The results show that in Taiwan’s base case,decreasing the hydraulic conductivity of the rock or decreasing the hydraulic conductivity of dikes results in a shorter transport path for sulfide and an increase in corrosion depth.However,the estimated canister failure time is still over one million years in the variant cases.
文摘<正>The disposal of spent nuclear fuel is a long-standing issue in nuclear technology.Mainly,UO_2 and metallic U arc used as a fuel in nuclear reactors.Spent nuclear fuel contains fission products and transuranium elements,which would remain radioactive for 10~4 to 10~8 years.In this brief communication,essential concepts and engineering elements related to high-level nuclear waste disposal are described.Conceptual design models are described and discussed considering the long-time scale activity of spent nuclear fuel or high level waste.Notions of physical and chemical barriers to contain nuclear waste are highiightened.Concerns regarding integrity,self-irradiation induced decomposition and thermal effects of decay heat on the spent nuclear fuel are also discussed.The question of retrievability of spent nuclear fuel after disposal is considered.
文摘To aid the United States Nuclear Regulatory Commission,Sandia National Laboratories (SNL) was contracted to investigate the seismic behavior of typical dry cask storage systems. Parametric evaluations characterized the sensitivity of calculated cask response characteristics to input parameters. The parametric evaluation investigated two generic cask designs (cylindrical and rectangular),three different foundation types (soft soil,hard soil,and rock),and three different casks to pad coefficients of friction (0.2,0.55,0.8) for earthquakes with peak ground accelerations of 0.25g,0.6g,1.0g and 1.25g. A total of 1 165 analyses were completed,with regression analyses being performed on the resulting cask response data to determine relationships relating the mean (16 % and 84 % confidence intervals on the mean) to peak ground acceleration,peak ground velocity,and pseudo-spectral acceleration at 1 Hz and 5 % damping. In general,the cylindrical casks experienced significantly larger responses in comparison to the rectangular cask. The cylindrical cask experienced larger top of cask displacements,larger cask rotations (rocking),and more occurrences of cask toppling (the rectangular cask never toppled over). The cylindrical cask was also susceptible to rolling once rocking had been initiated,a behavior not observed in the rectangular cask. Cask response was not overly sensitive to foundation type,but was significantly dependent on the response spectrum employed.
文摘Experience of operating reactor facilities (RF) with lead-bismuth coolant (LBC) has revealed that it is possible to perform safe refueling in short terms if the whole core is replaced and a kit of the special refueling equipment is used. However, comparing with RFs of nuclear submarines (NS), in which at the moment of performance of refueling the residual heat release is small, at RF SVBR-100 in a month after the reactor has been shut down, at the moment of performance of refueling the residual heat release is about 500 kW. Therefore, it is required to place the spent removable unit (SRU) with spent fuel subassemblies (SFSA) into the temporal storage tank (TST) filled with liquid LBC, in which the conditions for coolant natural circulation (NC) and heat removal via the tank vessel to the water cooling system are provided. After the residual heat release has been lowered to the level allowing transportation of the TST with SRU in the transporting-package container (TPC), it is proposed to consider a variant of TPCs transportation to the special site. On that site after the SRU has been reloaded into the long storage tank (LST) filled with quickly solidifying liquid lead, the TPCs can be stored during the necessary period. Thus, the controlled storage of LSTs is realized during several decades untill the time when SNF reprocessing and NFC closing are becoming economically expedient. On that storage, the four safety barriers are formed on the way of the release of radioactive products into the environment, namely: fuel matrix, fuel element cladding, solid lead and steel casing of the LST.
基金the High-Performance Computing Center of Nanjing Tech University for supporting the computational resources
文摘Dry storage containers must be secured and reliable during long-term storage,and the effect of decay heat released from the internal spent fuel on the cask has become an important research topic.In this paper,a 3D computational fluid dynamics model is presented,and the accuracy of the calculation is verified,with computational errors of less than 6.2%.The thermal stress of the dry storage cask was estimated by coupling it with a transient temperature field.The total power remained constant and adjusting the power ratio of the inner and outer zones had a small effect on the stress results,with a maximum equivalent stress of approximately 5.2 kPa,which occurred at the lower edge of the shell.In the case of tilt,the temperature gradient varied in a wavy distribution,and the wave crest moved from right to left.Altering the tilt angle affects the air distribution in the annular gap,leading to the shell temperature being transformed,with a maximum equivalent stress of 202 MPa at the bottom of the shell.However,the equivalent stress in both cases was less than the yield stress(205 MPa).
基金Supported by China Postdoctoral Science Foundation Funded Project (20070420527)National Natural Science Foundation of China (10535050)
文摘Reactor neutrino oscillation experiments, such as Daya Bay, Double Chooz and RENO are designed to determine the neutrino mixing angle θ13 with a sensitivity of 0.01--0.03 in sin^2 2θ13 at 90% confidence level, an improvement over the current limit by more than one order of magnitude. The control of systematic uncertainties is critical to achieving the sin^22θ13 sensitivity goal of these experiments. Antineutrinos emitted from spent nuclear fuel (SNF) would distort the soft part of energy spectrum and may introduce a non-negligible systematic uncertainty. In this article, a detailed calculation of SNF neutrinos is performed taking account of the operation" of a typical reactor and the event rate in the detector is obtained. A further estimation shows that the event rate contribution of SNF neutrinos is less than 0.2% relative to the reactor neutrino signals. A global X2 analysis shows that this uncertainty will degrade the θ13 sensitivity at a negligible level.
基金Supported by National Natural Science Foundation of China (10890094)
文摘The Daya Bay Reactor Antineutrino Experiment is designed to determine the as yet unknown neutrino mixing angle,θ13,by measuring the disappearance of electron antineutrinos from several nuclear reactor cores.The projected sensitivity in sin2(2θ13) of better than 0.01 at a 90% CL should be achieved after three years of data-taking.Antineutrinos emitted from spent nuclear fuel (SNF) distort the soft part of the energy spectrum.In this article,a calculation of the antineutrino spectra from the long-life isotopes in SNF is performed.A non-equilibrium generation of long half-life isotopes during the running time of the reactor is also analyzed.Finally,we show that the antineutrino event rate contribution from SNF,which has been stored in the SNF pool for several years,may be non-negligible.
基金financially supported by the National Natural Science Foundation of China(Grant Nos.50974049 and 51174068)the Heilongjiang Postdoctoral Science Foundation of China(Grant No.LBH-Z09009)
文摘A new physical simulation model was developed to simulate the casting process of the ductile iron heavy section spent-nuclear-fuel container.In this physical simulation model,a heating unit with DR24 Fe-Cr-Al heating wires was used to compensate the heat loss across the non-natural surfaces of the sample,and a precise and reliable casting temperature controlling/monitoring system was employed to ensure the thermal behavior of the simulated casting to be similar to the actual casting.Also,a mould system was designed,in which changeable mould materials can be used for both the outside and inside moulds for different applications.The casting test was carried out with the designed mould and the cooling curves of central and edge points at different isothermal planes of the casting were obtained.Results show that for most isothermal planes,the temperature control system can keep the temperature differences within 6 ℃ between the edge points and the corresponding center points,indicating that this new physical simulation model has high simulation accuracy,and the mould developed can be used for optimization of casting parameters of spent-nuclear-fuel container,such as composition of ductile iron,the pouring temperature,the selection of mould material and design of cooling system.In addition,to maintain the spheroidalization of the ductile iron,the force-chilling should be used for the current physical simulation to ensure the solidification of casting in less than 2 h.
文摘High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (~22.5%), increase the energy output (~41%), decrease the quantity of spent fuels to be disposed (~2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.
文摘For an analysis of the oxidation behavior of UO2 nuclear fuel pellet under a loss of water coolant accident in a spent nuclear fuel pool of an LWR, thermodynamic assessments of UO2 oxidation were carried out under various atmospheric conditions. In a steam atmosphere, it was assessed that UO2 would not be fully oxidized into U3O8 due to the relatively lower oxygen partial pressure, while UO2 will be fully oxidized into U3O8 in an air atmosphere. In an air and steam mixture atmosphere, the UO2 oxidation was dominantly affected by the air volumetric fraction, because of the relatively higher oxygen partial pressure of air. In addition, the effect of H2 volumetric fraction on the oxygen partial pressure under a mixture atmosphere was calculated, and it was revealed that UO2 pellet oxidation could be reduced above the critical value of H2 volumetric fraction.