## 2 Materials and Methods
### Gamma Spectrometry Analysis and Evaluation of Radiological Hazard Effects  
In the laboratory, for the gamma spectrometry analysis, samples were crushed, sieved at 1 mm (for homogeneity) and kept moisture-free in an oven in order to reach a constant weight. They were packed in a 20 mL polyethylene plastic vial to reach geometric homogeneity around the detector; then, the respective net weights were measured and recorded.  
Subsequently they were counted for 70,000 s and spectra were analyzed in order to obtain the activity concentration of \({}^{137}\)Cs and \({}^{40}\)K through the evaluation of their \(\gamma\)-lines at 661.66 keV and 1460.81 keV, respectively.  
The experimental setup was composed by a positive biased Ortec HPGe detector (GEM) (FWHM of 1.85 keV; peak to Compton ratio of 64:1; relative efficiency of 40% at the 1.33 MeV \({}^{60}\)Co \(\gamma\)-line) placed inside lead wells to shield the background radiation environment. Efficiency and energy calibrations were performed using a multipeak Marinelli geometry gamma source (AK-5901) of 1 L capacity, covering the energy range 59.54 keV-1836 keV and customized to reproduce the exact geometries of  
\begin{table}
\begin{tabular}{l c c c} \hline \hline
**No** & **Sample ID** & **Sample Name** & **County of Origin** \\ \hline
1 & S1 & Cyprus black & Cyprus island \\
2 & S2 & Himalayan pink & Himalayas \\
3 & S3 & Hawaii red & Hawaii islands \\
4 & S4 & iodized & Italy \\
5 & S5 & Hyposodic iodized & Italy \\
6 & S6 & Maldon smoked sea & England \\
7 & S7 & Common sea & Italy \\
8 & S8 & Breton sea & England \\
9 & S9 & Persia blue & Iran \\ \hline \hline \end{tabular}
\end{table}
Table 1: Investigated samples name, together their ID code and country of origin.  
samples in a water-equivalent epoxy resin matrix. The ANGLE 4 code was employed for the efficiency transfer factors calculations to the 20 mL vial sample holder geometry [20]. The Gamma Vision (Ortec) software was used for data acquisition and analysis [21].  
A photo of the gamma spectrometry experimental setup is reported in Figure 2a.  
The activity concentration of investigated radioisotopes was calculated using the following formula [22]:  
\[C=\frac{N_{E}}{\varepsilon_{E}t\gamma_{d}M} \tag{1}\]  
where N\({}_{E}\) indicates the net area of the radioisotope photopeak, \(\varepsilon_{E}\) and \(\gamma_{d}\) are the efficiency and yield of the photopeak, M is the mass sample (g) and t is the live time (s).  
The measurement result uncertainty, coverage factor k = 2, was calculated taking into account the following components: uncertainty of the counting estimation, of the calibration source, of the efficiency calibration, of the background subtraction and of the \(\gamma\)-branching ratio [21].  
The evaluation of radiological hazard effects was made in terms of the annual effective dose for food salt ingestion. It was calculated by the following [23]:  
\[D_{ing}\left(\frac{\mathrm{Sv}}{\mathrm{y}}\right)=\ h_{ing,K\to 40}\times J_{ing,\ K -40} \tag{2}\]  
where \(h_{ing,K\to 40}\) is the coefficient of effective dose for the insertion unit, for the ingestion of \({}^{40}\)K (Sv/Bq) and \(J_{ing,K\to 40}\) is the intake of \({}^{40}\)K (Bq/year) [23]. The latter value is obtained by multiplying the annual salt food consumption for the activity concentration of the investigated radionuclide experimentally measured.