Introduction

Human exposure to environmental radiation comes from a number of naturally occurring and anthropogenic sources. Natural radioactive materials such as uranium, thorium, and 40K can be found everywhere in the world. In addition, humans are exposed to external sources radiations from cosmic and terrestrial radiation. According to United Nations Scientific Committee of the Effects of Atomic Radiation (UNSCEAR), the worldwide average effective dose is 2.42 mSv of which, 0.87 mSv comes from external gamma exposure and 1.55 mSv from internal exposure of inhalation and ingestion dose [1].

The determination of indoor radiation dose for human health is very important and thus, the knowledge of the natural activity level of construction materials such as cement is necessary [2]. The raw materials that are used in cement industry are obtained from soil and rock, which contain diverse amounts of radionuclides of uranium (238U) and thorium (232Th) series, and the radioactive isotope of potassium (40K). In the 238U series, the segment starting from radium (226Ra) is radiological the most important one and, therefore, it is commonly referred to as 226Ra instead of 238U [3]. It is well recognized that some construction materials are naturally more radioactive than others. The natural radioactive materials used in construction even at low level causes an increase external and internal indoor exposure [4]. The external radiation exposure is caused by gamma radiation of the uranium and thorium decay and from 40K. The internal radiation exposure, mainly affecting the respiratory tract, is due to the short-lived daughter products of radon, which are released from materials to the indoor air [5]. People spend about 80% of their life inside houses and workplaces. Therefore, determination of primordial radionuclides (226Ra, 232Th and 40K) and attenuation coefficient in building materials is important in order to protect humans from radiological hazards. Moreover, such knowledge is essential for the development of standards and guidelines for the use of these materials [6, 7].

In this study, the activities of natural radionuclides of 226Ra, 232Th and 40K, and radiological parameters such as radium equivalent activity, absorbed dose rate, gamma and external-internal hazard indices were determined to assess the radiation risk associated with the cement produced in Turkey.

Material and method

Sample collection and preparation

In this study, in order to measure the natural radioactivity levels of Portland cement (PC), CEM I type samples produced in 16 locations were selected representing all geographical regions in Turkey were selected. All PC samples were taken from local markets as 50 kg paper bags. Figure 1 shows the locations of 16 PC sample sources that are produced by Turkish cement plants.

Fig. 1
figure 1

Locations of sampling points

Full size image

The characteristics of CEM I which include clinker in range of 95–100 and 0–5% minor additive of Gypsum used in buildings are given in TS EN 197-1 [8] standard. This standard is a descriptive standard which includes the composition, properties and eligibility criteria of general cement grades. This standard was revised in Europe in November 2011 and in Turkey in February 2012 [8].

The natural radioactivity levels of cement samples were measured using gamma-ray spectrometry with an HPGe detector. All measurements were duplicated and data sets were given as mean values. All of the cement samples were used without any processing such as crushing, homogenization, and sieving since they were already in a powder form. Of each sample of cement, 1 kg was picked and dried in a temperature-controlled furnace at 110 °C for 20–24 h to completely remove moisture. After drying process, cement samples were cooled in the moisture-free atmosphere and transferred to 250 cm3 polyethylene beaker. Before counting, the beakers were completely hermetically sealed for 4 weeks to reach the secular equilibrium, where the rate of decay of the radon daughters becomes equal to that of the parent. This step was necessary to ensure that radon gas was confined within the volume and the daughters remained in the sample [3, 9].

Gamma spectrometric measurements

Gamma spectrometric measurements for each sample’s activity concentration were conducted with Canberra Inc. manufactured with Extended Range Coaxial high-purity germanium (HPGe) detector placed in Canberra Standard Lead Shield, Canberra DSA1000 digital spectrum analyzer, Genie2000 Basic Spectroscopy Software with Gamma, Interactive Peak Fit and QA/QC options. Properties of the HPGe detector were provided by the manufacturer as Relative efficiency: 50%, Crystal diameter (mm): 69.5, Crystal length (mm): 51.0, End-cap diameter (inch): 3.5, End-cap material: Al.

Energy and efficiency calibrations of the system were performed by using multinuclide gamma standard sources including 210Pb, 241Am, 109Cd, 57Co, 139Ce, 203Hg, 113Sn, 85Sr, 137Cs, 88Y and 60Co gamma rays (which covers the energy range between 46 and 1836 keV) supplied from Isotope Products Laboratories, Eckert & Ziegler, Berlin, Germany. For self-absorption effects to be taken into account, an epoxy source with 0.4, 1.0 g/cm3 density, and a sand source with 1.7 g/cm3 density were used. 170 ml polyethylene cylindrical beakers, which were also supplied to Isotope Products Laboratories for reference source filling, were used for measurements. Source to the detector distance was kept at approximately 0.5 cm and true coincidence corrections for the calibration and measurement of radionuclides of interest were performed using Gespecor software.

All sample counts and background counts were taken approximately for 1 day. The specific activity of 40K was measured directly by its own gamma-ray at 1460.8 keV while activities of 226Ra and 232Th were calculated from the weighted mean of their respective daughter products in secular equilibrium in proportion to their errors. The specific activity of 226Ra was calculated using 214Pb (295 and 352 keV) and 214Bi (609 and 1765 keV) while the specific activity of 232Th was calculated using 228Ac (338 and 911 keV) and 208Tl (583 keV).

The minimum detectable activities (MDAs) were calculated using Eq. 1 [10]:

$${\text{MDA}} = \frac{2.71 + 4.65 \sqrt B }{{\left( {\varepsilon \times t \times \gamma \times m} \right)}}$$
(1)

where B is the background counts, ε is the counting efficiency, γ is the gamma emission probability for the gamma of interest, m is the mass of sample and t is the measurement time. The MDA values were estimated to be between 0.7 and 1.8 Bq kg−1 for 226Ra, 0.7 and 2.0 Bq kg−1 for 232Th and 14.0 and 22.8 Bq kg−1 for 40K.

Radium equivalent activity (Raeq)

In order to compare the activity concentrations of 226Ra, 232Th, and 40K in studied cement samples on a combined scale, Radium equivalent activity (Raeq) levels were calculated from Eq. (2):

$${\text{Ra}}_{\text{eq}} = A_{\text{Ra}} + 1. 4 3A_{\text{Th}} + 0.0 7 7A_{\text{K}}$$
(2)

where, A Ra, A Th, and A K are the activity concentrations of 226Ra, 232Th, and 40K respectively, in Bq kg−1.

Criterion formula (CF)

A very useful comparison criterion formula (CF) was suggested by Ravinsankar et al. [11] for limiting the annual radiation dose from building materials. CF is given by Eq. (3):

$${\text{CF}} = \frac{{A_{\text{Ra}} }}{{740 \;{\text{Bq kg}}^{ - 1} }} + \frac{{A_{\text{Th}} }}{{520 \;{\text{Bq kg}}^{ - 1} }} + \frac{{A_{\text{K}} }}{{9620 \;{\text{Bq kg}}^{ - 1} }} < 1$$
(3)

where, A Ra, A Th, and A K are the activity concentrations of 226Ra, 232Th, and 40K respectively, in Bq kg−1.

Representative level index (RLI)

The level of gamma radioactivity associated with different concentrations of certain specific radionuclides can be estimated as a representative level index (RLI) [11, 12] as given by Eq. (4):

$${\text{RLI}} = \frac{1}{150}A_{\text{Ra}} + \frac{1}{100}A_{\text{Th}} + \frac{1}{1500}A_{\text{K}}$$
(4)

where, A Ra, A Th, and A K are the activity concentrations of 226Ra, 232Th, and 40K respectively, in Bq kg−1.

Estimation of the absorbed gamma dose rate (D R ) and annual effective dose rate (H R )

Absorbed gamma dose rate (D R ) in the indoor air, and annual effective dose rate (H R ) related to gamma-ray emission from the radionuclides (226Ra, 232Th, and 40K) in the building materials were assessed based on the data and the equations provided by UNSCEAR [1] and the European Commission (EC) [13]. Also, the dose conversion coefficients for the center of a standard room were calculated according to UNSCEAR and EC regulations. The dimensions of the standard room were 4 m × 5 m × 2.8 m. The thickness of the concrete walls, floors, ceiling, and the density of the building were 20 cm and 2350 kg/m−3, respectively. Absorbed gamma dose rate (D R ) was calculated from Eq. (5):

$$D_{R} \left( {{\text{nGy h}}^{ - 1} } \right) = 0. 9 2A_{\text{Ra}} + 1. 1A_{\text{Th}} + 0.0 8 { }A_{\text{K}}$$
(5)

where, A Ra, A Th, and A K are the activity concentrations of 226Ra, 232Th, and 40K respectively, in Bq kg−1. According to UNSCEAR [1], taking into account of the external exposure rates from terrestrial gamma radiation, population-weighted average value of worldwide absorbed dose rate in the air was found to be 84.00 nGy h−1. This value was adopted as a reference for comparison of D R parameter in this study.

Annual effective dose rate (H R ) was calculated from Eq. (6) as adopted from UNSCEAR with Ravisankar et al. coefficients’ arrangements [1, 11]:

$$H_{R} \left( {{\text{mSv year}}^{ - 1} } \right) = \, D_{R} \times { 8766 } \times \, 0. 2 { } \times \, 0. 7 { } \times { 1}0^{ - 6} \, = \, D_{R} \times \, 0.00 1 2 3$$
(6)

where, D R (nGy h−1) is given by Eq. (5). The worldwide average value of H R is 0.48 mSv year−1 which is adopted from UNSCEAR [1] as a reference value.

Gamma Index (I γ )

The gamma activity concentration index, I γ , was defined by the European Commission [13], and Righi and Bruzzi [14] and is given in Eq. (7):

$$I_{\gamma } = \frac{{A_{\text{Ra}} }}{{300 \;{\text{Bq kg}}^{ - 1} }} + \frac{{A_{\text{Th}} }}{{200\; {\text{Bq kg}}^{ - 1} }} + \frac{{A_{\text{K}} }}{{3000 \;{\text{Bq kg}}^{ - 1} }}$$
(7)

where, A Ra, A Th, and A K are the activity concentrations of 226Ra, 232Th, and 40K respectively, in Bq kg−1.

Radiation hazard indices

Radiation hazard indices were measured using the two hazard indices as defined by Beretka and Mathew [15]. The first one was the external radiation hazard, H ex, and the latter was the internal radiation hazard, H in. Those indices are discussed below.

External radiation hazard (H ex)

External radiation hazard, H ex, has been adopted as a limit value of unity in order to be non-hazardous. It is given in Eq. (8):

$$H_{\text{ex}} = \frac{{A_{\text{Ra}} }}{{370 \;{\text{Bq kg}}^{ - 1} }} + \frac{{A_{\text{Th}} }}{{259 \;{\text{Bq kg}}^{ - 1} }} + \frac{{A_{\text{K}} }}{{4810\; {\text{Bq kg}}^{ - 1} }}$$
(8)

where, A Ra, A Th, and A K are the activity concentrations of 226Ra, 232Th, and 40K respectively, in Bq kg−1.

Internal radiation hazard (H in)

Internal radiation hazard, H in, is given in Eq. (9). The criterion value is less than or equal to unity [16].

$$H_{\text{in}} = \frac{{A_{\text{Ra}} }}{{185 \;{\text{Bq kg}}^{ - 1} }} + \frac{{A_{\text{Th}} }}{{259\; {\text{Bq kg}}^{ - 1} }} + \frac{{A_{\text{K}} }}{{4810 \;{\text{Bq kg}}^{ - 1} }}$$
(9)

where, A Ra, A Th, and A K are the activity concentrations of 226Ra, 232Th, and 40K respectively, in Bq kg−1.

Results and discussion

Activity concentration

The measured activity concentration of 226Ra, 232Th and 40K of the studied cement is given in Table 1. The lowest value of 226Ra activity of the analyzed cement samples was measured in sample PC16 (14.30 Bq kg−1), while the highest value for the same radionuclide was measured in PC05 (76.0 Bq kg−1, Table 1). The highest activity value for 232Th was still found in PC05 (34.0 Bq kg−1) and the lowest was 6.70 Bq kg−1 in PC11. The lowest value for 40K was 60.0 Bq kg−1 measured in PC06, and the highest value was 530.0 Bq kg−1 in PC08.

Table 1 Activity concentrations and some of the related parameter values in studied Portland cement (PC) samples of Turkey
Full size table

The contents of 226Ra, 232Th and 40K in tested cement samples depends on the raw materials and probably vary considerably in relation to the various geological source and geochemical characteristics. The mean activity concentrations of radionuclide parts in PCs were measured as 33.0, 16.7 and 239.5 Bq kg−1 for 226Ra, 232Th and 40K, respectively.

Radium equivalent activity (Raeq)

The activity concentrations of 226Ra, 232Th, and 40K determined in this study for PCs, were also compared with other studies shown in Table 2. In addition to this, the mean activity concentration values of 226Ra, 232Th, and 40K found to be 68, 80, and 45%, respectively, which were lower than the cement samples given in Table 2. On the other hand, the maximum activity concentration values of singular cement samples for of 226Ra, 232Th, and 40K, respectively, were lower than 11, 25, and 5% of cement samples given in Table 2.

Table 2 Comparison of natural radioactivity of Portland cements
Full size table

Raeq levels are given in Tables 1 and 2. All studied cement samples were compared with the specification limit of 370 Bq kg−1 [12]. Raeq values of cements in this study were between 36.0 and 144.9 Bq kg−1. The mean value of Raeq was calculated as 75.4 Bq kg−1. All Raeq values of measured PC are lower than the criterion limit of 370 Bq kg−1. Raeq values in Table 2 also shows that even in other PC samples around the world, Raeq is lower than the criterion limit value. However, mean Raeq value of cements in this study was 75.4 Bq kg−1 which was lower than 74% as shown in Table 2. Meanwhile, the maximum level of Raeq was 144.9 Bq kg−1, 17% lower compared to other cement samples given in Table 2.

Criterion formula (CF)

CF values of the cement samples were between 0.05 and 0.20 (mean value of CF = 0.10) and all values were considerably lower compared to the recommended maximum value 1.0. Therefore, it can be concluded from CF results that, the annual radiation dose from building materials manufactured using Turkish PC have minimal effect on human health.

Activity concentration comparison

Comparison of natural radioactivity of PCs and data sets of this study for 226Ra, 232Th and 40K is shown in Table 2. Same type comparison was also fulfilled for Radium-equivalent activity (Raeq) values for the same two sets of data.

Representative level index (RLI)

RLI values of PC samples of the present study were between 0.27 and 1.02. Mean value of RLI was 0.55. Except for PC05, all of the studied cement samples were lower than limit value of RLI, 1.00. However, RLI level of PC05 was only 1.02, which was very close to the limit. RLI level of PC05 exceed only 2% rate to the limit value 1.0, which was suggested by Alam et al. [49].

Estimation of the absorbed gamma dose rate (D R ) and annual effective dose rate (H R )

In this study, data set for absorbed gamma dose rates (D R ) in the indoor air for cement samples varied between 32.7 and 128.4 nGy h−1. The mean value of D R was 68.0 nGy h−1. Apart from 5 samples (PC02, PC04, PC05, PC08, and PC14), all the cement samples had lower values compared to the world average of D R  = 84.00 nGy h−1. In addition, the studied cement sample overall mean value of D R was lower than the world average reference value of 84.00 nGy h−1 (Table 1). Annual effective dose rate (H R ) values varied between 0.04 and 0.16 mSv year−1 (mean is 0.08 mSv year−1) for the studied cements. All the H R values were considerably lower than the average reference value of the world.

Gamma index (I γ )

Gamma index values of the studied cement samples are given in Table 3. Their minimum value was 0.13, while the maximum value was 0.51 with a mean of 0.27.

Table 3 Gamma index and radiation hazard values of studies Portland cement (PC) samples
Full size table

Radiation hazard indices

Two hazard indices, the external radiation hazard, H ex, and the internal radiation hazard, H in were studied.

External radiation hazard (H ex)

According to Table 3, the minimum, maximum, and mean values of H ex are 0.10, 0.39, and 0.20, respectively. These values were significantly less than unity, which was adopted as a reference limit value.

Internal radiation hazard (H in)

Table 3 shows that the minimum value, maximum value, and mean of H in were 0.14, 0.60, and 0.29, respectively. The limit values were acceptable, since they were less than a unity. Although sample PC05 gave a maximum value of 0.60, which was close to unity, the other values were generally low.

Conclusions

The activity concentrations of PC samples from different regions of Turkey were measured by gamma-ray spectrometry to determine the exposure of natural radiation and radiological parameters. The mean values of activity concentrations of 226Ra, 232Th, and 40K were 68, 80, and 45% respectively lower in comparison with the other cement samples as given in Table 2. The mean value of Raeq was found to be 75.4 Bq kg−1. All Raeq values of the measured PC samples were lower than the criterion limit of 370 Bq kg−1. The mean absorbed gamma dose rates (D R ) for the studied cement samples, was 68.0 nGy h−1 and this value was found to be lower than the average literature value of D R  = 84.00 nGy h−1.

This study assembles the radioactive data of cement samples from different regions of Turkey and it could be a handy database for future estimations of the impact of radioactive pollution as well as for the improving of the specific requirements for cements in the Turkey eco-labeling process. This work is also useful in maintaining radioactive data within the recommended ranges, especially RLI, D R , and keeping these parameters within the limits.