Most of the developed countries have used their tunnels as protective structures （public nuclear shelters） in case of nuclear emergencies to protect their citizens from the dangerous effects of nuclear weapons. The ...Most of the developed countries have used their tunnels as protective structures （public nuclear shelters） in case of nuclear emergencies to protect their citizens from the dangerous effects of nuclear weapons. The research aims to explain how to use tunnels to protect some people from neutrons and gamma rays and account the required safe depth. The computer code （MCNP5） is used at this model for such tunnel to account attenuation of both neutrons and gamma rays passing through the canal water, sand, soil and reinforced concrete wall layers. The last one （thickness 105 cm） constructed the tunnel construction. Also, the computer code is used to account the dose inside the tunnel, and account （neutron） dose, （neutron, gamma） dose, prompt （gamma） dose, total （gamma） dose and total （neutron ＋ gamma） dose estimated by μsv/h, at different depths from the water surface level （depths 200 cm, 500 cm, 1,000 cm, 1,600 cm, 2,200 cm, 2,600 cm, 2,800 cm, 3,400 cm, 3,700 cm, 4,000 cm and 4,600 cm）. Then, account these doses for three bombs （its intensity 20 KT, 100 KT and 1,000 KT）.展开更多
In the last few years, interest in burnup calculations using Monte Carlo methods has increased. Previous burnup codes have used diffusion theory for the neutronic portion of the codes. Diffusion theory works well for ...In the last few years, interest in burnup calculations using Monte Carlo methods has increased. Previous burnup codes have used diffusion theory for the neutronic portion of the codes. Diffusion theory works well for most reactors. However, diffusion theory does not produce accurate results in burnup problems that include strong absorbers or large voids. MCNPX code based on Mont Carlo Method, is used to design a three dimensional model for a BWR fuel assembly in a typical operating temperature and pressure conditions. A test case was compared with a benchmark problem and good agreement was found. The model is used to calculate the distribution of pin by pin power and flux inside the assembly. The effect of axial variation of water （coolant） density, and of control rods motion on the neutron flux and power distribution is analyzed. The effect of addition of Gd2O3 to natural uranium （0.711%） on both the thermal neutron flux and normalized power are analyzed. The concentration of U^235, U^238, Pu^239, and its isotopes is also calculated at burn-up 50 GWD/T.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Most of the developed countries have used their tunnels as protective structures （public nuclear shelters） in case of nuclear emergencies to protect their citizens from the dangerous effects of nuclear weapons. The research aims to explain how to use tunnels to protect some people from neutrons and gamma rays and account the required safe depth. The computer code （MCNP5） is used at this model for such tunnel to account attenuation of both neutrons and gamma rays passing through the canal water, sand, soil and reinforced concrete wall layers. The last one （thickness 105 cm） constructed the tunnel construction. Also, the computer code is used to account the dose inside the tunnel, and account （neutron） dose, （neutron, gamma） dose, prompt （gamma） dose, total （gamma） dose and total （neutron ＋ gamma） dose estimated by μsv/h, at different depths from the water surface level （depths 200 cm, 500 cm, 1,000 cm, 1,600 cm, 2,200 cm, 2,600 cm, 2,800 cm, 3,400 cm, 3,700 cm, 4,000 cm and 4,600 cm）. Then, account these doses for three bombs （its intensity 20 KT, 100 KT and 1,000 KT）.
文摘In the last few years, interest in burnup calculations using Monte Carlo methods has increased. Previous burnup codes have used diffusion theory for the neutronic portion of the codes. Diffusion theory works well for most reactors. However, diffusion theory does not produce accurate results in burnup problems that include strong absorbers or large voids. MCNPX code based on Mont Carlo Method, is used to design a three dimensional model for a BWR fuel assembly in a typical operating temperature and pressure conditions. A test case was compared with a benchmark problem and good agreement was found. The model is used to calculate the distribution of pin by pin power and flux inside the assembly. The effect of axial variation of water （coolant） density, and of control rods motion on the neutron flux and power distribution is analyzed. The effect of addition of Gd2O3 to natural uranium （0.711%） on both the thermal neutron flux and normalized power are analyzed. The concentration of U^235, U^238, Pu^239, and its isotopes is also calculated at burn-up 50 GWD/T.
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).
