2.1. Study Area
Samples of building materials, namely Brick, Sand, Gravel, Cement and Plaster, were collected from the city of Al-Shatra, DhiQar Governorate, Iraq. Al-Shatra is located 40 kilometers north of Nasiriyah, between latitudes 46° 11' 34' and 46° 8' 42' N. and longitude 31°23' 45' and 31° 26' 16' E. It is bordered from north by Al-Qadisiyah and Wasit Government, Al-Basra to the south, Maysan to the east, and Muthanna to the west. It is separated into five administrative divisions: Nasiriyah, Governorate center, Al-Rifai, Al-Shuyoukh Market, Al-Shatra and Al-Chibayish. Al-Shatra has an estimated area of 384 square kilometers, and a population of 354,000, according to the 2020 census [12].
2.2. Sample collection and preparation
From direct sales locations, fifteen samples of five distinct building materials were collected for measurements from Al-Shatra, Dhi-Qar Governorate, Iraq.The samples were dehydrated by placing them in an oven at 110°C for 24 hours, then crushed to approximately 2 mm thick particles, then each sample with 250 gm weight was stored in a closed Marinelli container for 30 days to let the growth of 222Rn and its decomposition products. Containers sent to physics lab at the University of Babylon, college of science, for natural radionuclide estimation by NaI(TI) (3"×3") gamma-ray spectrometry.
2.3. Gamma-ray spectrometry and statistical analysis
Gamma analysis was performed using a sodium iodide detector NaI (TI) (3"×3") inlay with thallium from the company (Alpha Spectra, Inc.-12I12/3) and a multi-channel analyzer (ORTEC-Digi Base) with a 4096-channel unit. Using standard sources 60Co, 22Na and 137Cs, an energy calibration for this detector was carried out. At the 662 keV photo peak of 137Cs, the detector's energy resolution is around 7.5%.The detector was shielded by a cylindrical lead shield to achieve the lowest radiation background. The measured spectra were analyzed by the MAESTRO32 software.
Statistical software (SPSS, version 21) was utilized for statistical analysis of the results and calculating p-value, which represents the probability of a coincidence in the effect of variables on each other. Also, the ANOVA and Tukey HSD tests were conducted to compare the means of the variables and determine whether the results are statistically significant or not.
2.4. Theoretical calculations
Specific activity of 226Ra, 232Th, and 40K can be calculated using the following equation [11].
$$\varvec{A} (\varvec{B}\varvec{q}/\varvec{k}\varvec{g})=\frac{\varvec{N}}{\varvec{\epsilon }. \varvec{I}\varvec{\gamma }. \varvec{m}. \varvec{t}} \left(1\right)$$
Where, A represents the sample's particular radionuclide activity, N denotes the net area under photo-peak (c/s), Ԑis the detector's efficiency at specific gamma energy, Iγ is the gamma-ray transition probability, m is the weight of sample (kg), and t is the time (sec) for the spectrum to be recorded.
Radium equivalent activity(Ra eq ) which is employed to evaluate the hazard of the concentration due to the effectiveness of 226Ra, 232Th, and 40K, and it can be estimated according to Eq. 2, where ARa, ATh, and AK are the activity concentrations of aforementioned radionuclides, respectively [11, 13].
$${\varvec{R}\varvec{a}}_{\varvec{e}\varvec{q}}(\varvec{B}\varvec{q}/\varvec{k}\varvec{g}) = {\varvec{A}}_{\varvec{R}\varvec{a}}+ 1.43 {\varvec{A} }_{\varvec{T}\varvec{h}}+ 0.077 {\varvec{A}}_{\varvec{K}} \left(2\right)$$
Internal (H in ) and external hazard indices (H ex ) can be measured using the following formulas [13, 14].
$${\varvec{H}}_{\varvec{i}\varvec{n}}=\frac{{\varvec{A}}_{\varvec{R}\varvec{a}}}{370}+\frac{{\varvec{A}}_{\varvec{T}\varvec{h}}}{259}+\frac{{\varvec{A}}_{\varvec{K}}}{4810}<1 \left(3\right)$$
$${\varvec{H}}_{\varvec{e}\varvec{x}}=\frac{{\varvec{A}}_{\varvec{R}\varvec{a}}}{185}+\frac{{\varvec{A}}_{\varvec{T}\varvec{h}}}{259}+\frac{{\varvec{A}}_{\varvec{K}}}{4810}<1 \left(4\right)$$
Activity concentration index (Iγ)of the construction material samples was used to assess the amount of gamma radiation risk related to natural gamma emitters in the studied material. The activity concentration index can estimated as follows [15]:
$${\varvec{I}}_{\gamma }=\frac{{\varvec{A}}_{\varvec{R}\varvec{a}}}{150}+\frac{{\varvec{A}}_{\varvec{T}\varvec{h}}}{100}+\frac{{\varvec{A}}_{\varvec{K}}}{1500} \left(5\right)$$
Indoor air absorbed dose rate (D) because of gamma ray emissions from naturally occurring radioactive elements in construction materials is calculated using Eq. 6 [13, 16].
$$\varvec{D} \left(\varvec{n}\varvec{G}\varvec{y}/\varvec{h}\right)=0.92{\varvec{A}}_{\varvec{R}\varvec{a}}+1.1{\varvec{A}}_{\varvec{T}\varvec{h}}+0.08{\varvec{A}}_{\varvec{k}} \left(6\right)$$
Annual effective dose may be derived using a conversion factor of 0.7 SvGy–1, which is employed with indoor and outdoor occupancy of 80% and 20%, respectively, to convert the absorbed rate to human effective dose equivalent. The following formulas are used to calculate the annual effective doses [16]:
$$\varvec{A}\varvec{E}\varvec{D}\left(\varvec{i}\varvec{n}\right)\left(\varvec{m}\varvec{S}\varvec{v}/\varvec{y}\right)=\varvec{D}(\varvec{n}\varvec{G}\varvec{y}/\varvec{h})\times {10}^{-6}\times 8760(\varvec{h}/\varvec{y})\times 0.8\times 0.7(\varvec{S}\varvec{v}/\varvec{G}\varvec{y}\left) \right(7)$$
$$\varvec{A}\varvec{E}\varvec{D}\left(\varvec{o}\varvec{u}\varvec{t}\right) \left(\varvec{m}\varvec{S}\varvec{v}/\varvec{y}\right)=\varvec{D}(\varvec{n}\varvec{G}\varvec{y}/\varvec{h})\times {10}^{-6}\times 8760(\varvec{h}/\varvec{y})\times 0.2\times 0.7(\varvec{S}\varvec{v}/\varvec{G}\varvec{y}\left) \right(8)$$
Exposure rate of gamma-ray (Ẋ) in air at one meter above a thick, massively stretched slab that is evenly dispersed throughout the material is provided by Eq. 9 [11, 16].
$$\varvec{Ẋ} \left(\varvec{\mu }\varvec{R}/\varvec{h} \right)=1.90{\varvec{A}}_{\varvec{R}\varvec{a}}+2.82{\varvec{A}}_{\varvec{T}\varvec{h}}+0.19{\varvec{A}}_{\varvec{K}} \left(9\right)$$