Investigating the time-dependent behavior of nuclear reactors during loss of flow accidents is necessary for safety assessment.Coupled neutronic/thermal-hydraulic codes are used to simulate a full three-dimensional mo...Investigating the time-dependent behavior of nuclear reactors during loss of flow accidents is necessary for safety assessment.Coupled neutronic/thermal-hydraulic codes are used to simulate a full three-dimensional model and predict the essential safety parameters.MCNP6/ANSYS-FLUENT17.2 coupled scheme is used in the present study to simulate a three-dimensional model for VVER-1000 assembly and analyze its behavior during a LOFA(loss of flow accident).Three LOFA scenarios are proposed to represent the failure of one,two or three of the coolant pumps.The influence of the accident on the reactivity and axial power distribution of the assembly is determined considering thermal-hydraulic feedbacks.Then the data obtained are provided to the thermal-hydraulic code to calculate the actual temperature values.The results of the study showed that the developed coupling scheme granted an actual and precise description of the axial behavior of the assembly during LOFA.The output data obtained from both neutronic and thermal-hydraulic calculations have a strong feedback effect;this demonstrated the effect of data exchange between codes to predict accurate values for the main safety parameters.Moreover,it revealed the importance of studying the detailed axial distribution of the safety parameters for the reactor assessment during accidents rather than taking average values in calculations.展开更多
针对坐落于意大利帕维亚大学的TRIGA Mark Ⅱ反应堆热柱结构进行优化设计,从而满足面向硼中子俘获治疗(BNCT)的单光子发射计算机断层成像(SPECT)研究要求。为提高计算效率并减小统计误差,对比分析使用SSW/SSR方法与直接使用反应堆为源...针对坐落于意大利帕维亚大学的TRIGA Mark Ⅱ反应堆热柱结构进行优化设计,从而满足面向硼中子俘获治疗(BNCT)的单光子发射计算机断层成像(SPECT)研究要求。为提高计算效率并减小统计误差,对比分析使用SSW/SSR方法与直接使用反应堆为源项时热柱内照射位置处中子能谱,其结果基本一致,从而验证了SSW/SSR方法的可靠性。为在该反应堆开展BNCT中SPECT实验,热柱中子束需准直为笔形束。对比分析四种热柱优化方案下束流口处及探测器处热中子和光子通量:40cm长石墨(射束口5cm×3cm);0.5cm厚硼包裹40cm长石墨(射束口5cm×3cm);30cm长天然锂聚乙烯(射束口直径4cm);30cm长天然锂聚乙烯(20cm长射束口直径5cm,5cm长射束口直径4cm,5cm长射束口直径2cm)。结果显示,射束口处热中子通量分别为1.05×108,2.52×107,6.08×107和5.10×107#/(cm2·s)。综合考虑中子准直效果及光子污染,方案三具有最优性能。为后续进行BNCT-SPECT理论和实验研究提供了基础,从而有效促进BNCT剂量准确评估方法的研究进程。展开更多
金刚石材料具有优异的耐高温、抗辐照性能,用其制作的辐照探测器在反应堆等苛刻环境下具有很好的应用前景。在分析金刚石中子探测器的结构和工作原理的基础上,使用MCNP(Monte Carlo N Particle Transport Code)模拟程序构建了金刚石中...金刚石材料具有优异的耐高温、抗辐照性能,用其制作的辐照探测器在反应堆等苛刻环境下具有很好的应用前景。在分析金刚石中子探测器的结构和工作原理的基础上,使用MCNP(Monte Carlo N Particle Transport Code)模拟程序构建了金刚石中子探测器的物理模型,考虑探测器用于2 MWt液态燃料钍基熔盐试验堆(Thorium Molten Salt experimental Reactor-Liquid Fueled,TMSR-LF1)辐射场中,计算中子转换层(6LiF、10B)厚度、金刚石厚度、γ甄别阈值对探测器的中子探测效率、γ探测效率以及n/γ抑制比的影响。结果表明:6LiF更适合在中子、γ混合场中用作中子转换层;随着6LiF厚度增加,中子探测效率先增大后减小,6LiF的最优厚度为25μm;金刚石厚度增大会导致探测器的n/γ甄别性能下降,可以采用设置γ甄别阈值的方法解决金刚石层过厚时带来的γ干扰过大的问题,使探测器达到对γ不灵敏的要求。模拟研究工作获得了探测器结构参数对探测器性能的影响规律,对探测器后续的制作和研究具有指导意义。展开更多
AP1000是美国西屋公司研发的大型压水反应堆,采用先进的非能动安全系统。AP1000反应堆有两种堆芯燃料布置方案:D19和Adv。结合两种设计方案的优点提出了一种新的堆芯燃料布置方案。利用MCNP6(Monte Carlo N-particle 6)程序对D19堆芯和...AP1000是美国西屋公司研发的大型压水反应堆,采用先进的非能动安全系统。AP1000反应堆有两种堆芯燃料布置方案:D19和Adv。结合两种设计方案的优点提出了一种新的堆芯燃料布置方案。利用MCNP6(Monte Carlo N-particle 6)程序对D19堆芯和新方案堆芯的首循环进行建模,并主要计算了新堆芯的核设计参数随燃耗的变化。结果表明,新堆芯在首循环寿期内满足AP1000的主要核设计准则。通过大规模并行计算表明,带燃耗计算功能的蒙特卡罗程序MCNP6能够在堆芯设计工作中发挥很好的参考作用。展开更多
Nuclear facility aging is one of the biggest problems encountered in nuclear engineering. Radiation damage is among one of the aging causes. This kind of damage is an important factor of mechanical properties deterior...Nuclear facility aging is one of the biggest problems encountered in nuclear engineering. Radiation damage is among one of the aging causes. This kind of damage is an important factor of mechanical properties deterioration. The interest of this study is on the Es-Salam research reactor aluminum vessel aging due to neutron radiation. Monte Carlo(MC) simulations were performed by MCNP6 and SRIM codes to estimate the defects created by neutrons in the vessel. MC simulations by MCNP6 have been performed to determine the distribution of neutron fluence and primary knock-on atom(PKA) creation. Considering our boundary conditions of the calculations, the helium and hydrogen gas production in the model at a normalized total neutron flux of 6.62×10^(12) n/cm^2 s were determined to be 2.86 × 10~8 and 1.33 × 10~9 atoms/cm^3 s,respectively. The SRIM code was used for the simulation of defects creation(vacancies, voids) in the aluminum alloy of the Es-Salam vessel(EsAl) by helium and hydrogen with an approximate energy of 11 MeV each.The coupling between the two codes is based upon postprocessing of the particle track(PTRAC) output file generated by the MCNP6. A small program based on the Mat Lab language is performed to condition the output file MCNP6 in the format of a SRIM input file. The concentration of silicon was determined for the vessel by the calculation of the total rate of ^(27)Al(n,γ)^(28)Si reaction. The DPA(displacement per atom) was calculated in SRIM according to R.E. Stoller recommendations; the calculated value is 0.02 at a fast neutron fluence 1.89 × 10^(19) n/cm^2.RCC-MRx standard for 6061-T6 aluminum was used for the simulation of the evolution of mechanical properties for high fluence. The calculated values of nuclear parameters and DPA obtained were in agreement with the experimental results from the Oak Ridge High Flux Isotope Reactor(HFIR) reported by Farrell and coworkers.展开更多
文摘Investigating the time-dependent behavior of nuclear reactors during loss of flow accidents is necessary for safety assessment.Coupled neutronic/thermal-hydraulic codes are used to simulate a full three-dimensional model and predict the essential safety parameters.MCNP6/ANSYS-FLUENT17.2 coupled scheme is used in the present study to simulate a three-dimensional model for VVER-1000 assembly and analyze its behavior during a LOFA(loss of flow accident).Three LOFA scenarios are proposed to represent the failure of one,two or three of the coolant pumps.The influence of the accident on the reactivity and axial power distribution of the assembly is determined considering thermal-hydraulic feedbacks.Then the data obtained are provided to the thermal-hydraulic code to calculate the actual temperature values.The results of the study showed that the developed coupling scheme granted an actual and precise description of the axial behavior of the assembly during LOFA.The output data obtained from both neutronic and thermal-hydraulic calculations have a strong feedback effect;this demonstrated the effect of data exchange between codes to predict accurate values for the main safety parameters.Moreover,it revealed the importance of studying the detailed axial distribution of the safety parameters for the reactor assessment during accidents rather than taking average values in calculations.
文摘针对坐落于意大利帕维亚大学的TRIGA Mark Ⅱ反应堆热柱结构进行优化设计,从而满足面向硼中子俘获治疗(BNCT)的单光子发射计算机断层成像(SPECT)研究要求。为提高计算效率并减小统计误差,对比分析使用SSW/SSR方法与直接使用反应堆为源项时热柱内照射位置处中子能谱,其结果基本一致,从而验证了SSW/SSR方法的可靠性。为在该反应堆开展BNCT中SPECT实验,热柱中子束需准直为笔形束。对比分析四种热柱优化方案下束流口处及探测器处热中子和光子通量:40cm长石墨(射束口5cm×3cm);0.5cm厚硼包裹40cm长石墨(射束口5cm×3cm);30cm长天然锂聚乙烯(射束口直径4cm);30cm长天然锂聚乙烯(20cm长射束口直径5cm,5cm长射束口直径4cm,5cm长射束口直径2cm)。结果显示,射束口处热中子通量分别为1.05×108,2.52×107,6.08×107和5.10×107#/(cm2·s)。综合考虑中子准直效果及光子污染,方案三具有最优性能。为后续进行BNCT-SPECT理论和实验研究提供了基础,从而有效促进BNCT剂量准确评估方法的研究进程。
文摘金刚石材料具有优异的耐高温、抗辐照性能,用其制作的辐照探测器在反应堆等苛刻环境下具有很好的应用前景。在分析金刚石中子探测器的结构和工作原理的基础上,使用MCNP(Monte Carlo N Particle Transport Code)模拟程序构建了金刚石中子探测器的物理模型,考虑探测器用于2 MWt液态燃料钍基熔盐试验堆(Thorium Molten Salt experimental Reactor-Liquid Fueled,TMSR-LF1)辐射场中,计算中子转换层(6LiF、10B)厚度、金刚石厚度、γ甄别阈值对探测器的中子探测效率、γ探测效率以及n/γ抑制比的影响。结果表明:6LiF更适合在中子、γ混合场中用作中子转换层;随着6LiF厚度增加,中子探测效率先增大后减小,6LiF的最优厚度为25μm;金刚石厚度增大会导致探测器的n/γ甄别性能下降,可以采用设置γ甄别阈值的方法解决金刚石层过厚时带来的γ干扰过大的问题,使探测器达到对γ不灵敏的要求。模拟研究工作获得了探测器结构参数对探测器性能的影响规律,对探测器后续的制作和研究具有指导意义。
文摘AP1000是美国西屋公司研发的大型压水反应堆,采用先进的非能动安全系统。AP1000反应堆有两种堆芯燃料布置方案:D19和Adv。结合两种设计方案的优点提出了一种新的堆芯燃料布置方案。利用MCNP6(Monte Carlo N-particle 6)程序对D19堆芯和新方案堆芯的首循环进行建模,并主要计算了新堆芯的核设计参数随燃耗的变化。结果表明,新堆芯在首循环寿期内满足AP1000的主要核设计准则。通过大规模并行计算表明,带燃耗计算功能的蒙特卡罗程序MCNP6能够在堆芯设计工作中发挥很好的参考作用。
文摘Nuclear facility aging is one of the biggest problems encountered in nuclear engineering. Radiation damage is among one of the aging causes. This kind of damage is an important factor of mechanical properties deterioration. The interest of this study is on the Es-Salam research reactor aluminum vessel aging due to neutron radiation. Monte Carlo(MC) simulations were performed by MCNP6 and SRIM codes to estimate the defects created by neutrons in the vessel. MC simulations by MCNP6 have been performed to determine the distribution of neutron fluence and primary knock-on atom(PKA) creation. Considering our boundary conditions of the calculations, the helium and hydrogen gas production in the model at a normalized total neutron flux of 6.62×10^(12) n/cm^2 s were determined to be 2.86 × 10~8 and 1.33 × 10~9 atoms/cm^3 s,respectively. The SRIM code was used for the simulation of defects creation(vacancies, voids) in the aluminum alloy of the Es-Salam vessel(EsAl) by helium and hydrogen with an approximate energy of 11 MeV each.The coupling between the two codes is based upon postprocessing of the particle track(PTRAC) output file generated by the MCNP6. A small program based on the Mat Lab language is performed to condition the output file MCNP6 in the format of a SRIM input file. The concentration of silicon was determined for the vessel by the calculation of the total rate of ^(27)Al(n,γ)^(28)Si reaction. The DPA(displacement per atom) was calculated in SRIM according to R.E. Stoller recommendations; the calculated value is 0.02 at a fast neutron fluence 1.89 × 10^(19) n/cm^2.RCC-MRx standard for 6061-T6 aluminum was used for the simulation of the evolution of mechanical properties for high fluence. The calculated values of nuclear parameters and DPA obtained were in agreement with the experimental results from the Oak Ridge High Flux Isotope Reactor(HFIR) reported by Farrell and coworkers.
