## 2 Simulations of Radiation Contamination Using MCNP 6.2  
Monte Carlo N-particle (MCNP) is a general-purpose Monte-Carlo-based code developed and maintained at Los Alamos National Laboratory, used for particle transport  
\begin{table}
\begin{tabular}{c c c c} \hline
**Survey Technique** & **Autonomy** & **Advantages** & **Limitations** \\ \hline Uniform survey & Possible through pre-planned motion paths & Well-established standard approach & Takes considerably more time to survey large areas \\ \hline Non-uniform survey & Typically controlled by human operator & Increased accuracy/precision in area of interest & Dependent on human operations \\ \hline Multi-sensor uniform survey & Possible through pre-planned motion paths & Decrease in survey time compared to single-sensor survey & Prone to complexities in motion path planning \\ \hline Multi-sensor & Typically controlled by human operator & Improved coverage in area of interest compared to single-sensor survey & Requires multiple human operators \\ \hline \end{tabular}
\end{table}
Table 1: Summary of advantages and limitations of typical survey methods for radiological contamination mapping.  
simulations. The code allows users to establish application-specific geometries and source terms to simulate the expected outcomes from particle transport through various materials; quantities such as radiation dose rates can be calculated. In this work, we simulated radiation contamination by using MCNP6.2 [20] to create datasets for testing the proposed mapping technique. A one-square-kilometer area consisting of several distinct contamination locations with varying levels of activity was simulated. The three main contamination sites consisted of many large circular surface sources emitting mono-energetic gamma rays from the decay of radioactive Cs-137 (i.e., at 662 keV). Each circular source had an assumed surface depth of 1 cm, and the gamma rays were emitted isotropically with varying radii between 65 to 150 m. The spatial emission profile within each surface source is uniform and perfectly flat. The total activity for all three contamination sites were scaled to be roughly twice the background level. The gamma rays are transported through dry sea-level atmospheric air [21], and a Cartesian mesh tally is imposed on the one-square-kilometer area at a predefined survey altitude to assess the flux at any volume element of 5 m \(\times\) 5 m \(\times\) 5 m. For each volume element, MCNP6.2 calculates an average areal flux in units of \(\gamma\)/cm\({}^{2}\)-s, which is used as the effective areal sensitivity of a given detector (i.e., 100% efficiency). A flux-to-dose conversion is implemented through the MCNP6.2 built-in dose-conversion functions that utilize the ICRP 60 standards [20]. The ground-level material is soil with a depth of 30 cm; with an elemental composition of hydrogen (weight fraction of 2.4%), oxygen (weight fraction of 59.8%), aluminum (weight fraction of 8.0%), and silicon (weight fraction of 29.7%) [21]. Assuming an average activity of 0.59 Bq/g from the soil [22] and a soil density of 1.52 g/cm\({}^{3}\)[21], the total activity solely from the soil is approximately \(2.7\times 10^{11}\) Bq. Background radiation from the soil was simulated as random fluctuations by taking samples from a Gaussian distribution with a mean equivalent to the average activity per volume of approximately 0.90 Bq/cm\({}^{3}\) at ground level. We assume a survey altitude of 40 m in this work to demonstrate a typical flight altitude of rotor-copter-type UAVs. Figure 1 shows the simulated dose rate distribution within the area of interest at an altitude of 40 m.