Neutron-induced fission is an important research object in basic science.Moreover,its product yield data are an indispensable nuclear data basis in nuclear engineering and technology.The fission yield tensor decomposi...Neutron-induced fission is an important research object in basic science.Moreover,its product yield data are an indispensable nuclear data basis in nuclear engineering and technology.The fission yield tensor decomposition(FYTD)model has been developed and used to evaluate the independent fission product yield.In general,fission yield data are verified by the direct comparison of experimental and evaluated data.However,such direct comparison cannot reflect the impact of the evaluated data on application scenarios,such as reactor transport-burnup simulation.Therefore,this study applies the evaluated fission yield data in transport-burnup simulation to verify their accuracy and possibility of application.Herein,the evaluated yield data of235U and239Pu are applied in the transport-burnup simulation of a pressurized water reactor(PWR)and sodium-cooled fast reactor(SFR)for verification.During the reactor operation stage,the errors in pin-cell reactivity caused by the evaluated fission yield do not exceed 500 and 200 pcm for the PWR and SFR,respectively.The errors in decay heat and135Xe and149Sm concentrations during the short-term shutdown of the PWR are all less than 1%;the errors in decay heat and activity of the spent fuel of the PWR and SFR during the temporary storage stage are all less than 2%.For the PWR,the errors in important nuclide concentrations in spent fuel,such as90Sr,137Cs,85Kr,and99Tc,are all less than 6%,and a larger error of 37%is observed on129I.For the SFR,the concentration errors of ten important nuclides in spent fuel are all less than 16%.A comparison of various aspects reveals that the transport-burnup simulation results using the FYTD model evaluation have little difference compared with the reference results using ENDF/B-Ⅷ.0 data.This proves that the evaluation of the FYTD model may have application value in reactor physical analysis.展开更多
In this study,a numerical flow model of the fission products(FPs)in the primary loop system of a molten salt reactor(MSR)was established and solved using Mathematica 7.0.The simulation results were compared with those...In this study,a numerical flow model of the fission products(FPs)in the primary loop system of a molten salt reactor(MSR)was established and solved using Mathematica 7.0.The simulation results were compared with those of the ORIGEN-S program in the static burnup mode,and the deviation was found to be less than 10%,which indicates that the results are in good agreement.Furthermore,the FPs distribution in the primary loop system under normal operating conditions of the 2 MW MSR was quantitatively analyzed.In addition,the distribution phenomenon of the FPs under different flow rate conditions was studied.At the end of life,the FPs activity in the core region(including active region,and upper and lower plenum regions)accounted for 77.3%,and that in the hot leg #1,main pump,hot leg #2,heat exchanger,and cold leg region accounted for 1.2%,16.15%,0.99%,2.5%,and 1.9%,respectively,of the total FPs in the primary loop under normal operating conditions.The proportion of FPs in the core decreased with the increase in flow rate in the range of 2.24-22,400 cm^3 s^-1.The established analytical method and conclusions of this study can provide an important basis for radiation safety design of the primary loop,radioactive source management design,thermal-hydraulic safety analysis,and radiochemical analysis of FPs of 2 MW MSRs.展开更多
The transmutation of long-lived fission products through spallation induced by light nuclides was investi-gated for the purpose of determining the feasibility of this approach for long-lived fission products,in both e...The transmutation of long-lived fission products through spallation induced by light nuclides was investi-gated for the purpose of determining the feasibility of this approach for long-lived fission products,in both economic and environmental terms.The cross-section data were obtained from the TALYS Evaluated Nuclear Data Library(TENDL).A thick target model was used to study the consumption of the target isotopes in the transmutation process.The transmutation yield was calculated using the highest beam intensity available with the China initiative accelerator-driven system.It was found that the light nuclide-induced spallation reaction can significantly reduce the radio toxicity of the investigated long-lived fission products.Using the transmutation target made of elemental LLFP and the proton beam with an intensity of 5 mA,the consumption of 90 Sr,93 Zr,107 Pd,or 137 Cs can reach approximately 500 g per year.展开更多
Chrome-molybdenum steel(2·1/4Cr1Mo) is one of the main products of steam generation.The adsorption behaviors of radioactive fission products on2·1/4Cr1Mo surface are critical in the analysis of HTR-PM.Here,t...Chrome-molybdenum steel(2·1/4Cr1Mo) is one of the main products of steam generation.The adsorption behaviors of radioactive fission products on2·1/4Cr1Mo surface are critical in the analysis of HTR-PM.Here,the adsorption behavior of cesium,strontium,silver and iodine on 2·1/4Cr1Mo was investigated with first-principle calculations that the Ag and I atoms prefer to be adsorbed at the square hollow site of the face-centered cubic iron cell with a binding energy of about 1 and 3 eV,respectively.In contrast,Cs and Sr atoms are not adsorbed on the surface of the 2·1/4Cr1Mo.These results are again confirmed via analysis of charge density differences and the densities of state.Furthermore,the adsorption rates of these fission products show that only I and Ag have significant adsorption on the metal substrate.These adsorption results explain the amount of adsorbed radionuclides for an evaluation of nuclear safety in HTR-PM.These micro-pictures of the interaction between fission products and materials are a new and useful way to analyze the source term.展开更多
A burnup calculation was performed to analyze the Apr`es ORIENT process, which aims to create highlyvaluable elements from fission products separated from spent nuclear fuels. The basic idea is to use nuclear transmut...A burnup calculation was performed to analyze the Apr`es ORIENT process, which aims to create highlyvaluable elements from fission products separated from spent nuclear fuels. The basic idea is to use nuclear transmutation induced by a neutron capture reaction followed by a β-decay, thus changing the atomic number Z of a target element in fission products by 1 unit. LWR(PWR) and FBR(MONJU) were considered as the transmutation devices. High rates of creation were obtained in some cases of platinum group metals(44Ru by FBR,46 Pd by LWR) and rare earth(64Gd by LWR,66 Dy by FBR). Therefore, systems based on LWR and FBR have their own advantages depending on target elements. Furthermore, it was found that creation rates of even Z(= Z + 1) elements from odd Z ones were higher than the opposite cases. This creation rate of an element was interpreted in terms of "average 1-group neutron capture cross section of the corresponding target element σc Z defined in this work. General trends of the creation rate of an even(odd) Z element from the corresponding odd(even) Z one were found to be proportional to the 0.78th(0.63th) power of σc Z, however with noticeable dispersion. The difference in the powers in the above analysis was explained by the difference in the number of stable isotopes caused by the even-odd effect of Z.展开更多
It has been found that the pressure in the reactor coolant system (RCS) remains high in some severe accident sequences at the time of reactor vessel failure, with the risk of causing direct containment heating (DCH). ...It has been found that the pressure in the reactor coolant system (RCS) remains high in some severe accident sequences at the time of reactor vessel failure, with the risk of causing direct containment heating (DCH). Intentional depressurization is an effective accident management strategy to prevent DCH or to mitigate its consequences. Fission product behavior is affected by intentional depressurization, especially for inert gas and volatile fission product. Because the pressurizer power-operated relief valves (PORVs) are latched open, fission product will transport into the containment directly. This may cause larger radiological consequences in containment before reactor vessel failure. Four cases are selected, including the TMLB' base case and the opening one, two and three pressurizer PORVs. The results show that inert gas transports into containment more quickly when opening one and two PORVs, but more slowly when opening three PORVs; more volatile fission product deposit in containment and less in reactor coolant system (RCS) for intentional depressurization cases. When opening one PORV, the phenomenon of revaporization is strong in the RCS.展开更多
The spallation cross-section data for the long-lived fission products (LLFPs) are scarce but required for the design of accelerator driven systems. In this paper, the isospin dependent quantum molecular dynamics model...The spallation cross-section data for the long-lived fission products (LLFPs) are scarce but required for the design of accelerator driven systems. In this paper, the isospin dependent quantum molecular dynamics model and the statistical code GEMINI are applied to simulate deuteron-induced spallation in the energy region of GeV/nucleon. By comparing the calculations with the experimental data, the applicability of the model is verified. The model is then applied to simulate the spallation of 90Sr, 93Zr, 107Pd, and 137Cs induced by deuterons at 200, 500 and 1000 MeV/nucleon. The cross-sections of isotopes, the cross-sections of long-lived nuclei, and the reaction energy are presented. Using the above observables, the feasibility of LLFP transmutation by spallation is discussed.展开更多
Cs and I can migrate through fuel-cladding interfaces and accelerate the cladding corrosion process induced by the fuel-cladding chemical interaction.Cr coating has emerged as an important candidate for mitigating thi...Cs and I can migrate through fuel-cladding interfaces and accelerate the cladding corrosion process induced by the fuel-cladding chemical interaction.Cr coating has emerged as an important candidate for mitigating this chemical interaction.In this study,first-principles calculations were employed to investigate the diffusion behavior of Cs and I in the Cr bulk and grain boundaries to reveal the microscopic interaction mitigation mechanisms at the fuel-cladding interface.The interaction between these two fission products and the Cr coating were studied systematically,and the Cs and I temperature-dependent diffusion coefficients in Cr were obtained using Bocquet’s oversized solute-atom model and Le Claire’s nine-frequency model,respectively.The results showed that the Cs and I migration barriers were significantly lower than that of Cr,and the Cs and I diffusion coefficients were more than three orders of magnitude larger than the Cr self-diffusion coefficient within the temperature range of Generation-IV fast reactors(below 1000 K),demonstrating the strong penetration ability of Cs and I.Furthermore,Cs and I are more likely to diffuse along the grain boundary because of the generally low migration barriers,indicating that the grain boundary serves as a fast diffusion channel for Cs and I.展开更多
The phenomenology involved in severe accidents in nuclear reactors is highly complex.Currently,integrated analysis programs used for severe accident analysis heavily rely on custom empirical parameters,which introduce...The phenomenology involved in severe accidents in nuclear reactors is highly complex.Currently,integrated analysis programs used for severe accident analysis heavily rely on custom empirical parameters,which introduce considerable uncertainty.Therefore,in recent years,the field of severe accidents has shifted its focus toward applying uncertainty analysis methods to quantify uncertainty in safety assessment programs,known as“best estimate plus uncertainty(BEPU).”This approach aids in enhancing our comprehension of these programs and their further development and improvement.This study concentrates on a third-generation pressurized water reactor equipped with advanced active and passive mitigation strategies.Through an Integrated Severe Accident Analysis Program(ISAA),numerical modeling and uncertainty analysis were conducted on severe accidents resulting from large break loss of coolant accidents.Seventeen uncertainty parameters of the ISAA program were meticulously screened.Using Wilks'formula,the developed uncertainty program code,SAUP,was employed to carry out Latin hypercube sampling,while ISAA was employed to execute batch calculations.Statistical analysis was then conducted on two figures of merit,namely hydrogen generation and the release of fission products within the pressure vessel.Uncertainty calculations revealed that hydrogen production and the fraction of fission product released exhibited a normal distribution,ranging from 182.784 to 330.664 kg and from 15.6 to 84.3%,respectively.The ratio of hydrogen production to reactor thermal power fell within the range of 0.0578–0.105.A sensitivity analysis was performed for uncertain input parameters,revealing significant correlations between the failure temperature of the cladding oxide layer,maximum melt flow rate,size of the particulate debris,and porosity of the debris with both hydrogen generation and the release of fission products.展开更多
Calculation of the decay heat from the decay/buildup of radionuclides generated after nuclear fission is one of the highest priorities in the nuclear industry. These calculations become more important if they are made...Calculation of the decay heat from the decay/buildup of radionuclides generated after nuclear fission is one of the highest priorities in the nuclear industry. These calculations become more important if they are made together with the analysis of the most important isotopes affecting the decay heat. They are useful in designing the necessary nuclear safety for spent fuels, and their importance cannot be overlooked in the designs of transporting fuel storage containers as well as in the management of the radioactive waste generated. In this paper, by using MATLAB, the decay heat after the thermal fission of a U-235 nucleus was numerically calculated by solving linear differential equations for all the buildups/decays of the fission products. Also, the most contribution of radioactive isotopes to the decay heat was analyzed by using Microsoft Excel. The most influential isotopes were deduced in two ways;either by calculating the most influential isotopes at specific times, or by determining the largest influences in a cumulative manner. All required nuclear data such as decay constants their branching ratios, independent fission yield, and average α-, β-, and γ-energies released per disintegration of any nuclide, have been extracted from the latest version of the Evaluated Nuclear Data Files (ENDF) database ENDF/B-VIII.0. The two different methods used showed a difference in the contributing isotopes, which is logical for the difference in the method of calculation. The first method is suitable for instantaneous data while the second method is more suitable when there is a need to know the cumulative calculations. In sum, we can say that both methods complement each other, and neither of them can be dispensed with in the accurate calculations related to transportation and storage of spent fuel.展开更多
A tokamak fusion-fission hybrid reatcor is proposed as one of candidates for disposal ofthe long-lived actinides and fission product wastes and supply of future energy.To assess thefeasibility of transmutation of long...A tokamak fusion-fission hybrid reatcor is proposed as one of candidates for disposal ofthe long-lived actinides and fission product wastes and supply of future energy.To assess thefeasibility of transmutation of long-lived radiowastes using fusion-fission hybrid reactors,afusion core design is presented and several possible conceptual blankets are studied,for,re-spectively,actinides transmutation and fission product transmutation.The results show thatactinides and fission products may be effectively transmuted using the presented hybrid reac-tors.展开更多
基金the National Natural Science Foundation of China(Nos.11875328,12075327 and 12105170)the Key Laboratory of Nuclear Data foundation(No.JCKY2022201C157)+1 种基金the Fundamental Research Funds for the Central Universities,Sun Yat-sen University(No.22lgqb39)the Open Project of Guangxi Key Laboratory of Nuclear Physics and Nuclear Technology(No.NLK2020-02).
文摘Neutron-induced fission is an important research object in basic science.Moreover,its product yield data are an indispensable nuclear data basis in nuclear engineering and technology.The fission yield tensor decomposition(FYTD)model has been developed and used to evaluate the independent fission product yield.In general,fission yield data are verified by the direct comparison of experimental and evaluated data.However,such direct comparison cannot reflect the impact of the evaluated data on application scenarios,such as reactor transport-burnup simulation.Therefore,this study applies the evaluated fission yield data in transport-burnup simulation to verify their accuracy and possibility of application.Herein,the evaluated yield data of235U and239Pu are applied in the transport-burnup simulation of a pressurized water reactor(PWR)and sodium-cooled fast reactor(SFR)for verification.During the reactor operation stage,the errors in pin-cell reactivity caused by the evaluated fission yield do not exceed 500 and 200 pcm for the PWR and SFR,respectively.The errors in decay heat and135Xe and149Sm concentrations during the short-term shutdown of the PWR are all less than 1%;the errors in decay heat and activity of the spent fuel of the PWR and SFR during the temporary storage stage are all less than 2%.For the PWR,the errors in important nuclide concentrations in spent fuel,such as90Sr,137Cs,85Kr,and99Tc,are all less than 6%,and a larger error of 37%is observed on129I.For the SFR,the concentration errors of ten important nuclides in spent fuel are all less than 16%.A comparison of various aspects reveals that the transport-burnup simulation results using the FYTD model evaluation have little difference compared with the reference results using ENDF/B-Ⅷ.0 data.This proves that the evaluation of the FYTD model may have application value in reactor physical analysis.
基金supported by the Chinese Academy of Sciences TMSR Strategic Pioneer Science and Technology Project(No.XDA02010000)The Frontier Science Key Program of Chinese Academy of Sciences(No.QYZDY-SSW-JSC016)
文摘In this study,a numerical flow model of the fission products(FPs)in the primary loop system of a molten salt reactor(MSR)was established and solved using Mathematica 7.0.The simulation results were compared with those of the ORIGEN-S program in the static burnup mode,and the deviation was found to be less than 10%,which indicates that the results are in good agreement.Furthermore,the FPs distribution in the primary loop system under normal operating conditions of the 2 MW MSR was quantitatively analyzed.In addition,the distribution phenomenon of the FPs under different flow rate conditions was studied.At the end of life,the FPs activity in the core region(including active region,and upper and lower plenum regions)accounted for 77.3%,and that in the hot leg #1,main pump,hot leg #2,heat exchanger,and cold leg region accounted for 1.2%,16.15%,0.99%,2.5%,and 1.9%,respectively,of the total FPs in the primary loop under normal operating conditions.The proportion of FPs in the core decreased with the increase in flow rate in the range of 2.24-22,400 cm^3 s^-1.The established analytical method and conclusions of this study can provide an important basis for radiation safety design of the primary loop,radioactive source management design,thermal-hydraulic safety analysis,and radiochemical analysis of FPs of 2 MW MSRs.
基金supported by the National Natural Science Foundation of China(No.11875328).
文摘The transmutation of long-lived fission products through spallation induced by light nuclides was investi-gated for the purpose of determining the feasibility of this approach for long-lived fission products,in both economic and environmental terms.The cross-section data were obtained from the TALYS Evaluated Nuclear Data Library(TENDL).A thick target model was used to study the consumption of the target isotopes in the transmutation process.The transmutation yield was calculated using the highest beam intensity available with the China initiative accelerator-driven system.It was found that the light nuclide-induced spallation reaction can significantly reduce the radio toxicity of the investigated long-lived fission products.Using the transmutation target made of elemental LLFP and the proton beam with an intensity of 5 mA,the consumption of 90 Sr,93 Zr,107 Pd,or 137 Cs can reach approximately 500 g per year.
基金supported by the National Science and Technology Major Project of the Ministry of Science and Technology of China(No.ZX06901)
文摘Chrome-molybdenum steel(2·1/4Cr1Mo) is one of the main products of steam generation.The adsorption behaviors of radioactive fission products on2·1/4Cr1Mo surface are critical in the analysis of HTR-PM.Here,the adsorption behavior of cesium,strontium,silver and iodine on 2·1/4Cr1Mo was investigated with first-principle calculations that the Ag and I atoms prefer to be adsorbed at the square hollow site of the face-centered cubic iron cell with a binding energy of about 1 and 3 eV,respectively.In contrast,Cs and Sr atoms are not adsorbed on the surface of the 2·1/4Cr1Mo.These results are again confirmed via analysis of charge density differences and the densities of state.Furthermore,the adsorption rates of these fission products show that only I and Ag have significant adsorption on the metal substrate.These adsorption results explain the amount of adsorbed radionuclides for an evaluation of nuclear safety in HTR-PM.These micro-pictures of the interaction between fission products and materials are a new and useful way to analyze the source term.
文摘A burnup calculation was performed to analyze the Apr`es ORIENT process, which aims to create highlyvaluable elements from fission products separated from spent nuclear fuels. The basic idea is to use nuclear transmutation induced by a neutron capture reaction followed by a β-decay, thus changing the atomic number Z of a target element in fission products by 1 unit. LWR(PWR) and FBR(MONJU) were considered as the transmutation devices. High rates of creation were obtained in some cases of platinum group metals(44Ru by FBR,46 Pd by LWR) and rare earth(64Gd by LWR,66 Dy by FBR). Therefore, systems based on LWR and FBR have their own advantages depending on target elements. Furthermore, it was found that creation rates of even Z(= Z + 1) elements from odd Z ones were higher than the opposite cases. This creation rate of an element was interpreted in terms of "average 1-group neutron capture cross section of the corresponding target element σc Z defined in this work. General trends of the creation rate of an even(odd) Z element from the corresponding odd(even) Z one were found to be proportional to the 0.78th(0.63th) power of σc Z, however with noticeable dispersion. The difference in the powers in the above analysis was explained by the difference in the number of stable isotopes caused by the even-odd effect of Z.
基金Supported by National Basic Research Program of China(No.2009CB724301)
文摘It has been found that the pressure in the reactor coolant system (RCS) remains high in some severe accident sequences at the time of reactor vessel failure, with the risk of causing direct containment heating (DCH). Intentional depressurization is an effective accident management strategy to prevent DCH or to mitigate its consequences. Fission product behavior is affected by intentional depressurization, especially for inert gas and volatile fission product. Because the pressurizer power-operated relief valves (PORVs) are latched open, fission product will transport into the containment directly. This may cause larger radiological consequences in containment before reactor vessel failure. Four cases are selected, including the TMLB' base case and the opening one, two and three pressurizer PORVs. The results show that inert gas transports into containment more quickly when opening one and two PORVs, but more slowly when opening three PORVs; more volatile fission product deposit in containment and less in reactor coolant system (RCS) for intentional depressurization cases. When opening one PORV, the phenomenon of revaporization is strong in the RCS.
基金Supported by the National Natural Science Foundation of China(11875328)
文摘The spallation cross-section data for the long-lived fission products (LLFPs) are scarce but required for the design of accelerator driven systems. In this paper, the isospin dependent quantum molecular dynamics model and the statistical code GEMINI are applied to simulate deuteron-induced spallation in the energy region of GeV/nucleon. By comparing the calculations with the experimental data, the applicability of the model is verified. The model is then applied to simulate the spallation of 90Sr, 93Zr, 107Pd, and 137Cs induced by deuterons at 200, 500 and 1000 MeV/nucleon. The cross-sections of isotopes, the cross-sections of long-lived nuclei, and the reaction energy are presented. Using the above observables, the feasibility of LLFP transmutation by spallation is discussed.
基金the National Natural Science Foundation of China(No.12375282)the Key Laboratory of Computational Physical Sciences Project(Fudan University),Ministry of Education.
文摘Cs and I can migrate through fuel-cladding interfaces and accelerate the cladding corrosion process induced by the fuel-cladding chemical interaction.Cr coating has emerged as an important candidate for mitigating this chemical interaction.In this study,first-principles calculations were employed to investigate the diffusion behavior of Cs and I in the Cr bulk and grain boundaries to reveal the microscopic interaction mitigation mechanisms at the fuel-cladding interface.The interaction between these two fission products and the Cr coating were studied systematically,and the Cs and I temperature-dependent diffusion coefficients in Cr were obtained using Bocquet’s oversized solute-atom model and Le Claire’s nine-frequency model,respectively.The results showed that the Cs and I migration barriers were significantly lower than that of Cr,and the Cs and I diffusion coefficients were more than three orders of magnitude larger than the Cr self-diffusion coefficient within the temperature range of Generation-IV fast reactors(below 1000 K),demonstrating the strong penetration ability of Cs and I.Furthermore,Cs and I are more likely to diffuse along the grain boundary because of the generally low migration barriers,indicating that the grain boundary serves as a fast diffusion channel for Cs and I.
基金This work was supported financially by the National Natural Science Foundation of China(No.12375176).
文摘The phenomenology involved in severe accidents in nuclear reactors is highly complex.Currently,integrated analysis programs used for severe accident analysis heavily rely on custom empirical parameters,which introduce considerable uncertainty.Therefore,in recent years,the field of severe accidents has shifted its focus toward applying uncertainty analysis methods to quantify uncertainty in safety assessment programs,known as“best estimate plus uncertainty(BEPU).”This approach aids in enhancing our comprehension of these programs and their further development and improvement.This study concentrates on a third-generation pressurized water reactor equipped with advanced active and passive mitigation strategies.Through an Integrated Severe Accident Analysis Program(ISAA),numerical modeling and uncertainty analysis were conducted on severe accidents resulting from large break loss of coolant accidents.Seventeen uncertainty parameters of the ISAA program were meticulously screened.Using Wilks'formula,the developed uncertainty program code,SAUP,was employed to carry out Latin hypercube sampling,while ISAA was employed to execute batch calculations.Statistical analysis was then conducted on two figures of merit,namely hydrogen generation and the release of fission products within the pressure vessel.Uncertainty calculations revealed that hydrogen production and the fraction of fission product released exhibited a normal distribution,ranging from 182.784 to 330.664 kg and from 15.6 to 84.3%,respectively.The ratio of hydrogen production to reactor thermal power fell within the range of 0.0578–0.105.A sensitivity analysis was performed for uncertain input parameters,revealing significant correlations between the failure temperature of the cladding oxide layer,maximum melt flow rate,size of the particulate debris,and porosity of the debris with both hydrogen generation and the release of fission products.
文摘Calculation of the decay heat from the decay/buildup of radionuclides generated after nuclear fission is one of the highest priorities in the nuclear industry. These calculations become more important if they are made together with the analysis of the most important isotopes affecting the decay heat. They are useful in designing the necessary nuclear safety for spent fuels, and their importance cannot be overlooked in the designs of transporting fuel storage containers as well as in the management of the radioactive waste generated. In this paper, by using MATLAB, the decay heat after the thermal fission of a U-235 nucleus was numerically calculated by solving linear differential equations for all the buildups/decays of the fission products. Also, the most contribution of radioactive isotopes to the decay heat was analyzed by using Microsoft Excel. The most influential isotopes were deduced in two ways;either by calculating the most influential isotopes at specific times, or by determining the largest influences in a cumulative manner. All required nuclear data such as decay constants their branching ratios, independent fission yield, and average α-, β-, and γ-energies released per disintegration of any nuclide, have been extracted from the latest version of the Evaluated Nuclear Data Files (ENDF) database ENDF/B-VIII.0. The two different methods used showed a difference in the contributing isotopes, which is logical for the difference in the method of calculation. The first method is suitable for instantaneous data while the second method is more suitable when there is a need to know the cumulative calculations. In sum, we can say that both methods complement each other, and neither of them can be dispensed with in the accurate calculations related to transportation and storage of spent fuel.
基金the High Technology Research and Development Programme of china.
文摘A tokamak fusion-fission hybrid reatcor is proposed as one of candidates for disposal ofthe long-lived actinides and fission product wastes and supply of future energy.To assess thefeasibility of transmutation of long-lived radiowastes using fusion-fission hybrid reactors,afusion core design is presented and several possible conceptual blankets are studied,for,re-spectively,actinides transmutation and fission product transmutation.The results show thatactinides and fission products may be effectively transmuted using the presented hybrid reac-tors.