Safety systems, built on state-of-the-art technology, are essential for achieving acceptable levels of plant safety to minimize hazards to the reactor and the general public. The second shutdown system(SSS) as an engi...Safety systems, built on state-of-the-art technology, are essential for achieving acceptable levels of plant safety to minimize hazards to the reactor and the general public. The second shutdown system(SSS) as an engineered safety feature and a part of the reactor protection system(RPS) is a means for rapidly shutting down a nuclear reactor, keeping it in a subcritical state and serving as a backup to the first shutdown system(FSS). In this research, one SSS with two types of optimum chamber designs is proposed that take into account the main current characteristic features of the Tehran research reactor with improvements over earlier designs. They are based on a liquid neutron absorber injection that is preferably different, diverse, and independent from the FSS based on the rod drop mechanism. The major design characteristics of this SSS with two different chambers were investigated using MCNPX 2.6.0 code. The performed calculations showed that the designed SSS is a reliable shutdown system, assuring an appropriate shutdown margin and injection time, with no significant effects on the effective delayed neutron fraction while causing minimal variations to the core structure. Further, the reasonable financial cost and the prolongation of the operation cycle are additional advantages of this design.展开更多
One of the most important safety features of nuclear facilities is the shielding material used to protect the operating personnel from radiation exposure. The most common materials used in radiation shielding are conc...One of the most important safety features of nuclear facilities is the shielding material used to protect the operating personnel from radiation exposure. The most common materials used in radiation shielding are concretes. In this study, a Monte Carlo N-Particle eXtended code is used to calculate the gamma-ray attenuation coefficients and dose rates for a new concrete material composed of MnFe_2O_4 nanoparticles, which is then compared with the theoretical and experimental results obtained for a SiO_2 nanoparticle concrete material. According to the results, the average relative differences between the simulations and the theoretical and experimental results for the linear attenuation coefficient(l) in the SiO_2 nanoparticle materials are 6.4% and 5.5%, respectively. By increasing the SiO_2 content up to 1.5% and the temperature of MnFe_2O_4 up to 673 K, l is increased for all energies. In addition, the photon dose rate decreases up to 9.2% and3.7% for MnFe_2O_4 and SiO_2 for gamma-ray energies of0.511 and 1.274 MeV, respectively. Therefore, it was concluded that the addition of SiO_2 and MnFe_2O_4 nanoparticles to concrete improves its nuclear properties and could lead to it being more useful in radiation shielding.展开更多
X-rays are commonly used for inspecting semiconductors. However, excessive radiation dose could damage semiconductors. Therefore, unnecessary exposure needs to be reduced. The ray quality, which is influenced by the t...X-rays are commonly used for inspecting semiconductors. However, excessive radiation dose could damage semiconductors. Therefore, unnecessary exposure needs to be reduced. The ray quality, which is influenced by the tube voltage and filter, determines the exposure. We designed an X-ray tube for inspecting semiconductors with different target/filter combinations and calculated X-ray spectra using the MCNPX （Monte Carlo n-particle extended） code. The target material was W, and the filters were made of Mo, W, and Zr. The W/W combination showed the lowest flux. The MCNPX code can reduce the development time and cost of the target/filter combination for inspecting semiconductors.展开更多
MCNPX(Monte Carlo N-Particle Transport Code)computer code is used to design a model to CANDU(Canada Deuterium Uranium)reactor core and its shielding system.It is assumed that reactor core is fueled with natural uraniu...MCNPX(Monte Carlo N-Particle Transport Code)computer code is used to design a model to CANDU(Canada Deuterium Uranium)reactor core and its shielding system.It is assumed that reactor core is fueled with natural uranium.The core radiation sources are calculated which consider prompt neutrons,neutron induced gamma and prompt gamma radiations.The total neutron flux and dose rate are calculated along the shield and at outer shield points.The results indicated that the major dose rate component at outer shield points is due to neutron induced gamma dose rate(μSv/h).展开更多
文摘Safety systems, built on state-of-the-art technology, are essential for achieving acceptable levels of plant safety to minimize hazards to the reactor and the general public. The second shutdown system(SSS) as an engineered safety feature and a part of the reactor protection system(RPS) is a means for rapidly shutting down a nuclear reactor, keeping it in a subcritical state and serving as a backup to the first shutdown system(FSS). In this research, one SSS with two types of optimum chamber designs is proposed that take into account the main current characteristic features of the Tehran research reactor with improvements over earlier designs. They are based on a liquid neutron absorber injection that is preferably different, diverse, and independent from the FSS based on the rod drop mechanism. The major design characteristics of this SSS with two different chambers were investigated using MCNPX 2.6.0 code. The performed calculations showed that the designed SSS is a reliable shutdown system, assuring an appropriate shutdown margin and injection time, with no significant effects on the effective delayed neutron fraction while causing minimal variations to the core structure. Further, the reasonable financial cost and the prolongation of the operation cycle are additional advantages of this design.
文摘One of the most important safety features of nuclear facilities is the shielding material used to protect the operating personnel from radiation exposure. The most common materials used in radiation shielding are concretes. In this study, a Monte Carlo N-Particle eXtended code is used to calculate the gamma-ray attenuation coefficients and dose rates for a new concrete material composed of MnFe_2O_4 nanoparticles, which is then compared with the theoretical and experimental results obtained for a SiO_2 nanoparticle concrete material. According to the results, the average relative differences between the simulations and the theoretical and experimental results for the linear attenuation coefficient(l) in the SiO_2 nanoparticle materials are 6.4% and 5.5%, respectively. By increasing the SiO_2 content up to 1.5% and the temperature of MnFe_2O_4 up to 673 K, l is increased for all energies. In addition, the photon dose rate decreases up to 9.2% and3.7% for MnFe_2O_4 and SiO_2 for gamma-ray energies of0.511 and 1.274 MeV, respectively. Therefore, it was concluded that the addition of SiO_2 and MnFe_2O_4 nanoparticles to concrete improves its nuclear properties and could lead to it being more useful in radiation shielding.
文摘X-rays are commonly used for inspecting semiconductors. However, excessive radiation dose could damage semiconductors. Therefore, unnecessary exposure needs to be reduced. The ray quality, which is influenced by the tube voltage and filter, determines the exposure. We designed an X-ray tube for inspecting semiconductors with different target/filter combinations and calculated X-ray spectra using the MCNPX （Monte Carlo n-particle extended） code. The target material was W, and the filters were made of Mo, W, and Zr. The W/W combination showed the lowest flux. The MCNPX code can reduce the development time and cost of the target/filter combination for inspecting semiconductors.
文摘MCNPX(Monte Carlo N-Particle Transport Code)computer code is used to design a model to CANDU(Canada Deuterium Uranium)reactor core and its shielding system.It is assumed that reactor core is fueled with natural uranium.The core radiation sources are calculated which consider prompt neutrons,neutron induced gamma and prompt gamma radiations.The total neutron flux and dose rate are calculated along the shield and at outer shield points.The results indicated that the major dose rate component at outer shield points is due to neutron induced gamma dose rate(μSv/h).
