New Cf-252 neutron source shielding design based Monte Carlo simulation using material combination

Abstract

Neutron shielding has been a worldwide concern for decades and appropriate methods in nuclear reactors and associated facilities for shielding have been developed. Light materials such as paraffin and water cannot be used in neutron radiography due to structural and fire concerns, and a material in the solid form is needed for effective neutron shielding. Therefore, the present study developed a new neutron shielding design for fixed industrial radiography facilities with solid structures based on material combination. Different materials were investigated to find the most appropriate combination design to shield the spontaneous neutron emitted from a Cf-252 source (with an energy in the range from several keV up to 20 MeV). The combination of iron, graphite, boron (or borate materials), and lead in this order, respectively, were found to be the most appropriate shielding structure for an open Cf-252 source used in fixed industrial radiography. As iron is characterized by a high removal cross section, its use in shielding the californium spontaneous neutron source is the key outcome of the present study. These results were confirmed with the Monte Carlo simulation-based particle and heavy ion transport code system.

Monte Carlo optimum management of 241Am/Be disused sealed radioactive sources

“…5 , 6 , and 7 , the horizontal effective dose profiles around the DSRS package are displayed. From these figures, it is observed that the distance needed to reduce the effective dose rate up to the ALARA acceptable limit for the public area (2.5 μSv/h) is 84.30 cm 14 , 16 , 27 , 30 , 39 , 54 , 55 . However, it appears to be 108.00 cm on the graph displayed in Fig.…”

“…Monte Carlo methods also named MC techniques are usually acknowledged as a group of computational algorithms based on the principle of repetition of random sampling to estimate unknown parameters and that use random number generators to solve problems that are stochastics in nature. MC methods are effective in modeling complex situations and problems with a high degree of freedom that cannot be analytically solved, by performing random numbers and random experiments associated with a specific probability density function 27 , 32 , 39 . The application of MC to neutron transport is one of the most appropriate means to evaluate neutron interaction in a medium since the neutron interaction with matter (or a medium) is a cross-section-dependent function.…”

“…Since neutron emitters as 241 Am/Be have a great variation of stochastic characteristics, their domestic long-term storage should be considered for optimizing the dismantling and disposal method. Such complex operation should not be mistakenly implemented and one of the most powerful tools that can be used to evaluate the waste package container radiological safety is the Monte Carlo computational method 8 , 9 , 27 – 32 . In the present study, the optimum encapsulation of 241 Am/Be disused sources based on Monte Carlo simulations for the disposal of DSRSs has been suggested.…”

Crack Detection during Laser Metal Deposition by Infrared Monochrome Pyrometer

“…Radiographic testing is a defect detection method which utilizes rays, such as gamma rays, neutron rays, and X-rays, to pass through dense materials, such as metal components, in order to determine, based on the degree of ray attenuation, whether there are cracks, pores, and other defects [ 35 , 36 , 37 , 38 , 39 ]. Nevertheless, this detection method is harmful to the human body [ 40 , 41 ], as well as expensive and inapplicable to on-line monitoring. On the other hand, metal magnetic powder inspection can locate defects on welded parts by measuring the variations in the magnetic field caused by the faulty parts [ 42 , 43 , 44 ].…”