## 2 Radiation Shielding Simulation  
For shielding simulation, MCNP (Monte Carlo N-Particle transport code) distributed by RSICC (Radiation Safety Information Computational Center) of the US National Laboratory was used [16; 17; 18; 19; 20; 21]. MCNP is mainly used for simulating nuclear fission processes, but because it can simulate the interaction of neutrons and photons with electrons or combined electrons/neutrons/photons among different particles, it is used in radiation shielding, radiation dosimetry, radiography and medical physics. Broadly, the input file contains geometric, material, source and tally specifications pertinent to the model under investigation. This is achieved via cell, surface and data cards (and terms pertinent to variance reduction if desired) defined in the input file [22; 23; 24; 25]. Simulation of a complex environment with many variables, such as radiation shielding, is difficult to derive theoretical results with computer code that applies deterministic methods. Therefore, MCNP, which can derive results by statistically approaching the interaction of radioactive particles with other substances, was used [16]. In order to reduce the computational error to less than 0.05%, it was set to repeat the experiment a sufficient number of times, and for lead and tungsten, a dispersion reduction technique was used to obtain more accurate results. In addition, rather than applying the energy intensity emitted by the actual radioactive particles, we tried to investigate the trend by adjusting the energy level step by step in the low energy region. To complement the structure, the number of layered structures of the layered composite was changed to 2, 4, 8, 16, 32, and 64 layers, and the thickness of each layer and the interface width of the two materials constituting the composite were designed differently. At this time, a layered composite was constructed so that the two materials were alternately positioned. The structure used in the simulation is shown in Figure 1. The simulation was carried out in a cuboid space of 30 cm \(\times\) 30 cm \(\times\) 100 cm that was covered with 30 cm of lead on all sides, and the radiation source was positioned 10 cm above the center of the floor, and the distance between the energy source and the composite material was set to 20 cm. The thickness of the designed composite material was 20 cm, and it was placed in the center of the cuboid space, and the value of the linear attenuation coefficient was obtained using the radiation dose before and after passing through the composite material. For the matrix of the composite material applied to the simulation, polyethylene (PE), which is a representative general-purpose polymer and is also used as a neutron shielding material, was used. In this simulation, the purpose of this simulation is to investigate the change in shielding ability due to the structure rather than the shielding effect by the material, and to analyze the shielding ability of the cesium adsorption nonwoven fabric using this, so PE, which is easily accessible in general, was used. As an inorganic material forming a composite structure with PE, concrete, lead, copper, tungsten, etc. are applied as radiation shielding materials in large structures. Inorganic materials were also selected mainly for materials commonly used in shielding materials, and four materials with different shielding abilities were used to investigate the conditions of composite attenuation due to the composite structure. Since it was not possible to consider all the various types of composite materials in the simulation model, the structure was simplified and layered, fiber, and particle composite structures were applied to the modeling. As shown in Table 1, in 4 types of composites with a combination of PE/concrete, PE/copper, PE/lead, and PE/tungsten, the gamma ray energy is changed at 1, 10, 30, 50, 80, and 100 MeV in a low energy region.