SW for needs of design of bilogical shielding, which allows to build a sandwich from three different materials with different thickness. For this combination, the input to the MCNP computing program is created. There the neutron flux density is calculated in front of the shielding and behind the shielding. In the program, there are defined materials, that are useful for protection against slow, epidermal, fast and high energy neutrons. They are concerete, polyethylene, boron polyethylene, tungsten, lead and steel. If new materiál is needed, the library can be extended. The SW also allows to determine the dose rate for cases where the power of the neutron source is known. In addition to the influence of neutrons, SW considers the production of secondary radiation, which may worsen the radiation situation behind the shielding. The content is subject to business secrets (§ 17 to 20 of the Commercial Act); the data are modified to make them public.
64-bit application executable under the operation system Windows 7 and higher.
Non-exclusive licence is provided in terms of a licence agreement. The license for using MCNP6.2 is not included in the offered software.
Authors gratefully acknowledge financial support from the Technology Agency of the Czech Republic under project No. TJ01000184 – Analysis of activation detectors for use in systems with high energy neutrons.