The radioactive isotope 60Co is used in many applications and is typically produced in heavy water reactors.As most of the commercial reactors in operation are pressurized light water reactors(PWRs),the world supply o...The radioactive isotope 60Co is used in many applications and is typically produced in heavy water reactors.As most of the commercial reactors in operation are pressurized light water reactors(PWRs),the world supply of high level radioactive cobalt would be greatly increased if 60Co could be produced in them.Currently,60Co production in PWRs has not been extensively studied;for the 59Co(n,c)60Co reaction,the positioning of 59Co rods in the reactor determines the rate of production.This article primarily uses the models of 60Co production in Canadian CANDU power reactors and American boiling water reactors;based on relevant data from the pressurized water Daya Bay nuclear power plant,a PWR core model is constructed with the Monte Carlo N-Particle Transport Code;this model suggests changes to existing fuel assemblies to enhance 60Co production.In addition,the plug rods are replaced with 59Co rods in the improved fuel assemblies in the simulation model to calculate critical parameters including the effective multiplication factor,neutron flux density,and distribution of energy deposition.By considering different numbers of 59Co rods,the simulation indicates that different layout schemes have different impact levels,but the impact is not large.As a whole,the components with four 59Co rods have a small impact,and the parameters of the reactor remain almost unchanged when four 59Co rods replace the secondary neutron source.Therefore,in theory,the use of a PWR to produce 60Co is feasible.展开更多
In this study, based differential equations methods are used to solve equations because these methods are dependent on boundary value data more than other mathematical equations. We have calculated neutron flux, criti...In this study, based differential equations methods are used to solve equations because these methods are dependent on boundary value data more than other mathematical equations. We have calculated neutron flux, criticality and geometrical eigenvalue by using multi-group method and solving the neutron diffusion equation for finite and infinite cylindrical and spherical reactors in this study. For the calculation of the total neutron flux cross sections, we need the neutron diffusion equation. Thus, we have established the relationship between neuron flow and cross-section of neuron depending on neutron energy. Critical calculations have been made by comparing the results with MNCP (montecarlo n-partical) simulation methods. For necessary computer calculations, the programme, Wolfram-Matematica-7 has been used.展开更多
The China Advanced Research Reactor (CARR) is scheduled to be operated in the autumn of 2008.In this paper,we report preparations for installing the neutron radiography instrument (NRI) and for utilizing it efficientl...The China Advanced Research Reactor (CARR) is scheduled to be operated in the autumn of 2008.In this paper,we report preparations for installing the neutron radiography instrument (NRI) and for utilizing it efficiently. The 2-D relative neutron intensity profiles for the water-vapor two-phase flow inside the robe were obtained using the MCNP code without influence of y-ray and electronic-noise.The MCNP simulation of the 2-D neutron intensity profile for the water-vapor two-phase flow was demonstrated.The simulated 2-D neutron intensity profiles could be used as the benchmark data base by calibrating part of the data measured by the CARR-NRI.The 3-D objective images allow us to understand the flow pattern more clearly and it is reconstructed using the MATLAB through the threshold transformation techniques.And thus it is concluded that the MCNP code and the MATLAB are very useful for constructing the benchmark data base for the investigation of the water-vapor two-phase flow using the CARR-NRI.展开更多
A fusion-fission hybrid conceptual reactor is established. It consists of a DT neutron source and a spherical shell of depleted uranium and hydrogen lithium. The tritium production rate （TPR） distribution in the con...A fusion-fission hybrid conceptual reactor is established. It consists of a DT neutron source and a spherical shell of depleted uranium and hydrogen lithium. The tritium production rate （TPR） distribution in the conceptual reactor was measured by DT neutrons using two sets of lithium glass detectors with different thicknesses in the hole in the vertical direction with respect to the D＋ beam of the Cockcroft-Walton neutron generator in direct current mode. The measured TPR distribution is compared with the calculated results obtained by the three- dimensional Monte Carlo code MCNP5 and the ENDF/B-VI data file. The discrepancy between the measured and calculated values can be attributed to the neutron data library of the hydrogen lithium lack S（α, β） thermal scattering model, so we show that a special database of low-energy and thermal neutrons should be established in the physics design of fusion-fission hybrid reactors.展开更多
The Nigeria Research Reactor-1 (NIRR-1) is one of the Commercial Miniature Neutron Source Reactors (MNSRs) sited outside China and scheduled for conversion under the auspices of Reduced Enrichment for Research and Tes...The Nigeria Research Reactor-1 (NIRR-1) is one of the Commercial Miniature Neutron Source Reactors (MNSRs) sited outside China and scheduled for conversion under the auspices of Reduced Enrichment for Research and Test Reactors (RERTR) program. Since 2006, the reduction in the fuel enrichment of MSNR facilities from greater than 90% HEU cores to less than 20% LEU cores has been embarked upon. Consequently in this work, the physics parameters of three dispersion LEU fuels, which include U3Si, U3Si2, and U9Mo enriched to 19.75% were determined by the MCNP code to investigate their suitability for the conversion of NIRR-1 to LEU. The following reactor core physics parameters were computed for the LEU fuel options: clean cold core excess reactivity (ρex), control rod (CR) worth, shut down margin (SDM), neutron flux distributions in the irradiation channels and kinetics data (i.e. effective delayed neutron fraction, βeff?and prompt neutron lifetime, lf). Results are compared with experimental and calculated data of the current HEU core and indicate that it would be feasible to use any of the LEU options for the conversion of commercial MNSR in general and NIRR-1 in particular from HEU to LEU.展开更多
作为获国际认可的第四代核电站反应堆堆型之一的熔盐堆(Molten salt reactor,MSR),具有固有安全性高、经济性好、核资源可持续发展以及易于防止核扩散等优点。针对石墨-熔盐零功率堆的几何参数,利用蒙特卡罗计算程序MCNP5建立了物理计...作为获国际认可的第四代核电站反应堆堆型之一的熔盐堆(Molten salt reactor,MSR),具有固有安全性高、经济性好、核资源可持续发展以及易于防止核扩散等优点。针对石墨-熔盐零功率堆的几何参数,利用蒙特卡罗计算程序MCNP5建立了物理计算模型,计算临界情况下堆芯径向、轴向中子通量及增殖区厚度与Th-U转换比(Conversion ratio,CR)的关系。结果表明,(1)石墨-熔盐零功率堆堆芯中子通量密度分布较为平坦;(2)石墨-熔盐零功率堆反射层厚度和增殖区厚度在一定范围内,CR随反射层厚度或增殖区厚度的增加而增加,当超出该范围,CR不再随反射层厚度或增殖区厚度的增加而明显增加。展开更多
基金supported by the National Natural Science Foundation of China(Nos.11635005 and 11705058)the Fundamental Research Funds for the Central Universities(No.2018ZD10)
文摘The radioactive isotope 60Co is used in many applications and is typically produced in heavy water reactors.As most of the commercial reactors in operation are pressurized light water reactors(PWRs),the world supply of high level radioactive cobalt would be greatly increased if 60Co could be produced in them.Currently,60Co production in PWRs has not been extensively studied;for the 59Co(n,c)60Co reaction,the positioning of 59Co rods in the reactor determines the rate of production.This article primarily uses the models of 60Co production in Canadian CANDU power reactors and American boiling water reactors;based on relevant data from the pressurized water Daya Bay nuclear power plant,a PWR core model is constructed with the Monte Carlo N-Particle Transport Code;this model suggests changes to existing fuel assemblies to enhance 60Co production.In addition,the plug rods are replaced with 59Co rods in the improved fuel assemblies in the simulation model to calculate critical parameters including the effective multiplication factor,neutron flux density,and distribution of energy deposition.By considering different numbers of 59Co rods,the simulation indicates that different layout schemes have different impact levels,but the impact is not large.As a whole,the components with four 59Co rods have a small impact,and the parameters of the reactor remain almost unchanged when four 59Co rods replace the secondary neutron source.Therefore,in theory,the use of a PWR to produce 60Co is feasible.
文摘In this study, based differential equations methods are used to solve equations because these methods are dependent on boundary value data more than other mathematical equations. We have calculated neutron flux, criticality and geometrical eigenvalue by using multi-group method and solving the neutron diffusion equation for finite and infinite cylindrical and spherical reactors in this study. For the calculation of the total neutron flux cross sections, we need the neutron diffusion equation. Thus, we have established the relationship between neuron flow and cross-section of neuron depending on neutron energy. Critical calculations have been made by comparing the results with MNCP (montecarlo n-partical) simulation methods. For necessary computer calculations, the programme, Wolfram-Matematica-7 has been used.
基金Supported by National Natural Science Foundation of China (Grant No.50876080)
文摘The China Advanced Research Reactor (CARR) is scheduled to be operated in the autumn of 2008.In this paper,we report preparations for installing the neutron radiography instrument (NRI) and for utilizing it efficiently. The 2-D relative neutron intensity profiles for the water-vapor two-phase flow inside the robe were obtained using the MCNP code without influence of y-ray and electronic-noise.The MCNP simulation of the 2-D neutron intensity profile for the water-vapor two-phase flow was demonstrated.The simulated 2-D neutron intensity profiles could be used as the benchmark data base by calibrating part of the data measured by the CARR-NRI.The 3-D objective images allow us to understand the flow pattern more clearly and it is reconstructed using the MATLAB through the threshold transformation techniques.And thus it is concluded that the MCNP code and the MATLAB are very useful for constructing the benchmark data base for the investigation of the water-vapor two-phase flow using the CARR-NRI.
基金Supported by Chinese Special Project for ITER (2010GB111002)
文摘A fusion-fission hybrid conceptual reactor is established. It consists of a DT neutron source and a spherical shell of depleted uranium and hydrogen lithium. The tritium production rate （TPR） distribution in the conceptual reactor was measured by DT neutrons using two sets of lithium glass detectors with different thicknesses in the hole in the vertical direction with respect to the D＋ beam of the Cockcroft-Walton neutron generator in direct current mode. The measured TPR distribution is compared with the calculated results obtained by the three- dimensional Monte Carlo code MCNP5 and the ENDF/B-VI data file. The discrepancy between the measured and calculated values can be attributed to the neutron data library of the hydrogen lithium lack S（α, β） thermal scattering model, so we show that a special database of low-energy and thermal neutrons should be established in the physics design of fusion-fission hybrid reactors.
文摘The Nigeria Research Reactor-1 (NIRR-1) is one of the Commercial Miniature Neutron Source Reactors (MNSRs) sited outside China and scheduled for conversion under the auspices of Reduced Enrichment for Research and Test Reactors (RERTR) program. Since 2006, the reduction in the fuel enrichment of MSNR facilities from greater than 90% HEU cores to less than 20% LEU cores has been embarked upon. Consequently in this work, the physics parameters of three dispersion LEU fuels, which include U3Si, U3Si2, and U9Mo enriched to 19.75% were determined by the MCNP code to investigate their suitability for the conversion of NIRR-1 to LEU. The following reactor core physics parameters were computed for the LEU fuel options: clean cold core excess reactivity (ρex), control rod (CR) worth, shut down margin (SDM), neutron flux distributions in the irradiation channels and kinetics data (i.e. effective delayed neutron fraction, βeff?and prompt neutron lifetime, lf). Results are compared with experimental and calculated data of the current HEU core and indicate that it would be feasible to use any of the LEU options for the conversion of commercial MNSR in general and NIRR-1 in particular from HEU to LEU.
文摘作为获国际认可的第四代核电站反应堆堆型之一的熔盐堆(Molten salt reactor,MSR),具有固有安全性高、经济性好、核资源可持续发展以及易于防止核扩散等优点。针对石墨-熔盐零功率堆的几何参数,利用蒙特卡罗计算程序MCNP5建立了物理计算模型,计算临界情况下堆芯径向、轴向中子通量及增殖区厚度与Th-U转换比(Conversion ratio,CR)的关系。结果表明,(1)石墨-熔盐零功率堆堆芯中子通量密度分布较为平坦;(2)石墨-熔盐零功率堆反射层厚度和增殖区厚度在一定范围内,CR随反射层厚度或增殖区厚度的增加而增加,当超出该范围,CR不再随反射层厚度或增殖区厚度的增加而明显增加。
