|Year : 2013 | Volume
| Issue : 2 | Page : 71-77
Preliminary investigation of naturally occurring radionuclide in some six representative cement types commonly used in Cameroon as building material
MM Ndontchueng1, Eric Jilbert Mekongtso Nguelem1, A Simo2, RL Njinga2, JF Beyala2, D Kryeziu3
1 National Radiation Protection Agency of Cameroon, Yaounde; Department of Physics, Faculty of Science, University of Douala, Douala, Cameroon
2 National Radiation Protection Agency of Cameroon, Yaounde, Cameroon
3 Canberra Packard Central Europe GmbH, Wienersiedlung 6, A-2432 Schwadorf, Austria
|Date of Web Publication||14-Mar-2014|
Eric Jilbert Mekongtso Nguelem
Department of Physics, Faculty of Science, University of Douala, P.O. Box 24157, Douala
Source of Support: The authors are grateful for the support and technical cooperation provided by the National Radiation Protection Agency of Cameroon in granting access to the facilities to successfully complete this study., Conflict of Interest: None
The present study was aimed at the determination of the specific activity of naturally occurring radionuclides and evaluation of the radiological health hazards in 24 cement samples representing six cement types commonly used in Cameroon for building construction have been analyzed. A high purity germanium detector spectrometer was used for quantification of gamma emitting radionuclide in the cements to demonstrate the radiological health hazards. Terrestrial absorbed dose rate (D), annual effective dose, radium equivalent activity, external/internal hazard index, activity gamma and alpha index caused by gamma emitting natural radionuclide are determined from the obtained values of 226 Ra, 232 Th and 40 K. The calculated values of the absorbed dose rate and the indoor annual effective dose are slightly higher as compared to the recommended world-wide values. The details of the samples preparation procedure and the gamma-ray spectrometry technique are presented, together with the preliminary investigated results of specific activity of naturally occurring radionuclide chains for six representative cement type-analyzed in this current work.
Keywords: Broad energy germanium detector, cement, natural radioactivity, radiological health hazard parameters
|How to cite this article:|
Ndontchueng M M, Nguelem EJ, Simo A, Njinga R L, Beyala J F, Kryeziu D. Preliminary investigation of naturally occurring radionuclide in some six representative cement types commonly used in Cameroon as building material. Radiat Prot Environ 2013;36:71-7
|How to cite this URL:|
Ndontchueng M M, Nguelem EJ, Simo A, Njinga R L, Beyala J F, Kryeziu D. Preliminary investigation of naturally occurring radionuclide in some six representative cement types commonly used in Cameroon as building material. Radiat Prot Environ [serial online] 2013 [cited 2020 Jul 7];36:71-7. Available from: http://www.rpe.org.in/text.asp?2013/36/2/71/128871
| Introduction|| |
Gamma radiation emitted from naturally occurring radioisotopes, such as 40 K and the radionuclides from the 232 Th and 238 U series and their decay products existing at trace levels in all ground formations pose external radiation hazard. The main external source of irradiation to the human body are from 238 U, 235 U and 232 Th, the parents of the three natural decay series, called the uranium (U) series, the actinium series and the thorium (Th) series, respectively. These series each consists of many daughter products which are generated through successive decay of the parent radionuclides. In the three long-lived series, decay cascades produce radioactive daughter nuclides which eventually result in stable isotopes, 208 Pb, 207 Pb and 206 Pb respectively.
Natural uranium is a composite of the isotopes 238 U (99.28%), 234 U (0.0057%) and 235 U (0.72%). On a mass basis, there is far more 238 U than 235 U in a natural uranium sample and the activity ratio is approximately 21:1.  The behavior and distribution of these decay series radionuclides in the environment are based on their biogeochemistry the half-life (t½ ) and the nature of their surroundings. The naturally occurring radioisotopes are present in different concentrations in sediments reflecting the origin of the sediments that are used in production of cements. The activity concentrations of 226 Ra, 2l4 Bi, 214 Pb (from 238 U decay series), 228 Ac, 212 Bi, 212 Pb (from 232 Th decay series) and 40 K were measured using a high purity germanium detector spectrometer.
Studies conducted on natural radioactivity have shown that the presence of potassium ( 40 K) and other daughter radionuclides from Thorium ( 232 Th) and Uranium ( 238 U) decay series in various components in the environment result in radiation exposure of the global population.  The primordial radionuclides are predominant in almost all raw and produced materials widely used in the building industry including; cement, bricks, sands, tile, limestone, gypsum and those derived from rocks and soils. , Natural radiation in building materials related to external and internal exposure. The external exposure comes from direct terrestrial gamma-rays radiation and the internal exposure from inhalation of radioactive inert gas Radon ( 222 Rn, a daughter product of 226 Ra). The presence of these natural radionuclides in building raw materials depend on geological/geographical conditions and the geochemical characteristics of the materials themselves.
Cement is commonly used in construction of building both in urban and rural areas in Cameroon. Due to its high production rate and widely used by the population, the radiation originating from it deserves special attention as this work seeks to achieve. Some of the raw materials used in the production of cement includes; limestone (CaCo 3 ), shale ash and iron oxide which contain elements such as gypsum, silicates and aluminates that have ionization tendency. 
According to the rule that exposure should be "as low as reasonably achievable," the radium equivalent, the external hazard (H ex ) index, the absorbed dose and the annual effective dose were assessed and compared with results of other studies and with the world-wide average value in the United Nations Scientific Committee on the effects of Atomic Radiation report.  In this regard, the evaluation of natural radioactivity was determined in some cement types used in Cameroon focusing on the radiation risk assessment of the population and the radionuclides of interest. The distribution of natural gamma emitting radionuclides and their respective annual effective dose rates, produced by 40 K, 238 U, 226 Ra, 235 U and 232 Th, were determined.
This subject is important in environmental radiological protection since cements are widely used as building material. The activity concentration of 226 Ra, 232 Th and 40 K in the different type of cements were measured by means of a broad energy germanium detector (BEGe). The most important radiological parameters including the absorbed dose rate, annual effective dose, radium equivalent activity (Ra eq ), the H ex and internal hazard (H in ) index, activity gamma index (Iγ) and alpha index (Iα) were also determined and evaluated for each cement sample.
| Sampling and Sample Preparation|| |
A total of six main types of cement manufactures (a number of four samples for each cement type) and used in building construction in Cameroon were collected. For each of the represented cement types, four samples were taken to make a composite sample. These samples were used without any process of homogenization since they are in powdered form. The samples were oven-dried at 110°C for 24 h to ensure that moisture is completely removed. The samples were sieved, packed in a cylindrical geometry, labeled, weighed and hermetically sealed with a plastic tape. The samples containers used were selected in order to match the container geometry used for efficiency calibration. The container geometry used was then a cylindrical container of 200 ml. The sealed samples were then stored for 30 days to enable them attain a state of secular equilibrium with their short-lived progeny.
| Experimental|| |
After the in-growth period, each water sample was subjected to a low background gamma-ray spectrometer consisting of BEGe (BE6530) manufactured by Canberra Industries. As reported by the manufacturer it has a resolution of 0.5 keV at 5.9 keV of 55 Fe, 0.75 keV at 122 keV of 57 Co and 2.2 keV at 1332 keV of 60 Co, respectively. To prevent high background counts due to external radioactive sources, with the intention to reduce the counting time and improve the detection limit, the detector is placed in a low-level Canberra Model 747 lead shield having a thickness of 10 cm. The inner part of the lead shield is covered with copper to reduce KX-rays from lead. Furthermore, a multiport II multichannel analyzer was used to generate energy distributions of the radioactive samples. In order to obtain a statistically good computational net peak area, each sample was measured for 86,400 s. The background has been evaluated before running the samples and it was measured for 172,800 s.
The efficiency calibration files have been generated by means of using Canberra designed Laboratory Sourceless Object Counting System (LabSOCS) mathematical calibration software that incorporates the characterization information of a BEGe6530 high-pure germanium detector. When generating the efficiency calibration file, the LabSOCS calibration software is taking into account all parameters related to these measurements including dimensions of the counting geometries, physical and chemical compositions as well as the distance source-to-detector end-cap. To validate the accuracy of the LabSOCS mathematical efficiency calibration, some test have been conducted comparing the LabSOCS generated efficiency results with the empirical peak efficiency for a 60 Co-60 point source positioned at a distance of 25 cm from the detector end-cap. The calculated results were in good agreement showing that differences between mathematical and empirical peak efficiencies are within 3-5%. To avoid error due to extrapolating the curve, the calibration curve is plotted in dual mode with cross-over energy at 165.85 keV ( 139 Ce). A fourth order polynomial equation was the best fit for the lower and higher energy curve and the fitting equations are the following:
LabSOCS Genie 2000, Gamma Acquisition V.3.2.1 and Gamma Analysis Software, V.3.2.3 was used for data acquisition and analysis.  Following the sample analysis, the specific activity concentration for each identified radionuclide has been reported in a unit of Becquerel per kilogram using the Equation 3 below. Furthermore, the software is taking care to automatically check and perform the interference correction and calculate also the weighted mean for those radionuclide that emit more than one gamma ray. In addition, CANBERRA's patented Cascade Summing Correction algorithms allows us to correct the nuclide activities for losses or gains due to the presence of cascade summing effect related to these close counting geometries.
where: N s is the net counts of the radionuclide in the samples; N B is the net counts of radionuclide in the background; Pγi is the gamma emission probability (gamma yield); ε (E i ) is the peak efficiency of the detector at energy E i; t s is the sample counting time; t B is the background measuring time; M s is the mass of the sample (kg) and C is the cascade summing correction co-efficient.
Assuming a state of secular equilibrium of 238 U and 232 Th and their respective decay daughter products, the following relatively intense gamma-ray transitions were used to measure the activity concentrations for the above-mentioned radionuclides.
- 226 Ra concentration was calculated as a weighted mean of the activity concentrations of the gamma-rays of 214 Pb (295.1 keV, 351.9 keV), 214 Bi (609.3 keV and 1120.29 keV) and its specific gamma-ray at 186.2 keV. Interference correction due to the presence of 185.7 keV energy peak of 235 U has been taken into account and subtracted accordingly
- The gamma-ray photopeaks used for the determination of the 232 Th contents were 338.4 keV, 911.2 keV and 969.11 keV of 228 Ac and 238.6 keV of 212 Pb
- 40 K was directly determined by using 1460.8 (10.7%) gamma-ray.
| Assessment of Dose|| |
Absorbed dose rate in air (D)
In order to assess any radiological hazard, the exposure to radiation arising from radionuclides present in cement can be determined in terms of many parameters. A direct connection between radioactivity concentrations of natural radionuclides and their exposure rate is known as the absorbed dose rate in the air at 1 meter above the ground surface. The mean activity concentrations of 226 Ra (of the 238 U series), 232 Th and 40 K (Bq/kg) in the soil samples are used to calculate the absorbed dose rate given using the following formula provided by UNSCEAR  and European Commission.  UNSCEAR and the European Commissions have provided the dose conversion co-efficients for the standard room centre.
Where D is the absorbed dose rate in nGy/h, A Ra , A Th and A K are the activity concentration of 226 Ra ( 238 U), 232 Th and 40 K, respectively. The dose co-efficient in units of nGy/h per Bq/kg were taken from the UNSCEAR (2000) report ,, and determined by the Monte Carlo simulation using the model standard room.
Annual effective dose equivalent
The absorbed dose rate in air at 1 m above the ground surface does not directly provide the radiological risk to which an individual is exposed.  The absorbed dose can be considered in terms of the AEDE from indoor terrestrial gamma radiation which is converted from the absorbed dose by taking into account two factors, namely the conversion coefficient from absorbed dose in air to effective dose and the indoor occupancy factor. The AEDE can be estimated using the following formula: ,
The values of those parameters used in the UNSCEAR report (2000) are 0.7 Sv/Gy for the conversion co-efficient from absorbed dose in air to effective dose received by adults and 0.8 for the indoor occupancy factor. 
Due to a non-uniform distribution of natural radionuclides in the soil samples, the actual activity level of 226 Ra, 232 Th and 40 K in the samples can be evaluated by means of a common radiological index named the Ra eq . , It is the most widely used index to assess the radiation hazards and can be calculated using Equation 4 given by Beretka and Mathew (1985). This estimates that 370 Bq/kg of 226 Ra, 259 Bq/kg of 232 Th and 4810 Bq/kg of 40 K produce the same gamma-ray dose rate. 
Where A Ra, A Th and A K are the activity concentration of 226 Ra, 232 Th and 40 K in Bq/kg, respectively.
The permissible maximum value of the Ra eq is 370 Bq/kg , which corresponds to an effective dose of 1 mSv for the general public and to the radiation dose rate of 1.5 mGy/year. ,
H ex and H in
To limit the radiation exposure attributable to natural radionuclides in the samples to the permissible dose equivalent limit of 1 mSv/y, the H ex index based on a criterion have been introduced using a model proposed by Krieger  which is given by. ,
In order to keep the radiation hazard insignificant, the value of H ex index must not exceed the limit of unity. The maximum value of H ex equal to unity corresponds to the upper limit of Ra eq 370 Bq/kg , measured dimensions and calculated densities.
In addition to the H ex , radon and its short-lived products are also hazardous to the repository organs. To account for this threat the maximum permissible concentration for 226 Ra must be reduced to half of the normal limit (185 Bq/kg). The internal exposure to carcinogenic radon and its short-lived progeny is quantified by the H in index given by the expression. 
Activity concentration index (Iγ)
Due to more than one radionuclide contribute to the dose; it is practical to present investigation levels in the form of an activity index. The European Commission has proposed in their guidance document the induction of an activity concentration index used to assess safety requirement for buildings materials: 
where A Ra , A Th and A K are the thorium, radium and potassium activity concentrations (Bq/kg).
Values of index Iγ ≤ 0.5 corresponds to a dose rate criterion of 0.3 mSv/year, whereas Iγ corresponds to a criterion of 1 mSv/year. ,, Thus the material with Iγ >6 should be avoided to use as building material since these values correspond to the dose rates higher than 1 mSv/year which is highest then recommended values. ,
Due to radon inhalation originated from buildings materials.  The Iα was determined using the following formula:
Where A Ra is the specify activity concentration of 226 Ra assumed in equilibrium with 238 U.
The recommended exemption and upper of 226 Ra activity concentration in building materials are 100 and 200 Bq/kg, respectively as suggested by many countries in the world.  These considerations reflected in the Iα. The recommended upper limit activity concentration of 226 Ra is 200 Bq/kg, for which Iα= 1.
| Results and Discussions|| |
The activity concentrations of 226 Ra, 232 Th and 40 K and Ra eq in different cement types used in Cameroon have been calculated and presented below in [Table 1]. The comparison of the mean values of 226 Ra, 232 Th, 40 K and Ra eq activities in the cement samples collected in Cameroon with data from other countries are reported in [Table 2]. The radiation hazard indices in the investigated cement have been estimated and summarized in [Table 3].
|Table 1: Specific activity concentration of 226 Ra, 232 Th and 40 K and radium equivalent activity in the investigated cement samples used in Cameroon|
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|Table 2: Comparison of the mean values of 226 Ra, 232 Th, 40 K and Ra eq activities in the studied cement samples collected in Cameroon with data from other countries|
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|Table 3: Values of radiation hazard parameters in the investigated cement type|
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The uncertainty of the average values is quoted at 1 sigma standard deviation.
The specific activities of the natural radionuclides of 226 Ra, 232 Th and 40 K were determined by using gamma-ray spectroscopy in different cement types used in Cameroon. The mean activity concentrations together with their associated uncertainties reported at 1 sigma standard deviation are shown in [Table 1]. The specific activity concentrations in the investigated cement types used in Cameroon were found to varies from 16.6 ± 0.7 to 47.6 ± 1.4 Bq/kg for 226 Ra; 12.5 ± 0.8-32.4 ± 0.9 Bq/kg for 232 Th and "below the detection limit (BDL)" to 285.7 ± 11.9 Bq/kg for 40 K, respectively. Although the lowest activity concentration of 226 Ra was observed in CW1 cement sample the highest value was in CW5 cement sample. For 232 Th the lowest activity value was observed in CW2 cement sample and the highest value was obtained in cement sample referred as CW3. It is known that potassium is present in almost all geological and raw material. In the case of 40 K, cement sample (CW4) showed the highest activity value and was BDL in CW6 cement type. This was very obvious as a result to the fact that BEGe detector exhibits low background than the typical coaxial detector because it is more transparent to high energy cosmogenic background radiation that permeates above background levels in laboratories and high energy gamma from naturally occurring radioisotopes such as 40 K.  As can be seen from [Table 1], the specific activity values of 226 Ra, 232 Th and 40 K determined in cement varied from one sample to another. These variations in activity concentrations of 226 Ra, 232 Th and 40 K in the investigated cements types used in Cameroon may depend on the uranium, thorium and potassium content under the earth crust from where the raw materials for a particular brand of cement were obtained.
As per the UNSCEAR, the world's mean values of 226 Ra, 232 Th and 40 K specific activity are 32, 45 and 420 Bq/kg, respectively.  The activity concentrations of 232 Th and 40 K observed in this study were significantly lower compared to the world average values, whereas the mean concentration of 226 Ra values observed in CW3, CW4, CW5 and CW6 cement type were relatively higher than the recommended values.
The distribution of 226 Ra, 232 Th and 40 K in the studies cement types was not uniform. Due to this non-uniformity of natural radionuclides in this study, the Ra eq was calculated to compare the specific activity of the studied cement samples and the results are summarized in below [Table 1]. The calculated radium equivalent ranged between 53.6 ± 7.4 and 105.9 ± 10.9 Bq/kg. It was clearly shown that obtained values of Ra eq in the studied cement types are lower than the recommended maximum value of 370 Bq/kg, which corresponds to an annual effective dose of 1 mSv. This shows that investigated cement types are within the recommended safety limit when used as building construction materials.
For comparison purposes, the activity concentrations of 226 Ra, 232 Th, 40 K as well as the calculated radium equivalent have been compared with the data reported by other countries. As can be seen from [Table 2], some of the observed values of 226 Ra, 232 Th, 40 K and the evaluated radium equivalent were lower than the reported data of other countries while some were higher but varied in within the reported mean values of other countries.
The calculated H in and H ex indices for the investigated cement types as shown in [Table 3] varied from 0.2 to 0.4 and from 0.1 to 0.3, respectively. These calculated H in and H ex indices in all cement samples are indeed less than unity.
The Iγ and Iα used to assess safety requirement for building materials were evaluated and presented in [Table 3]. The obtained values for both of them ranged from 0.2 ± 0.0 to 0.4 ± 0.0 and from 0.1 to 0.2 respectively. The obtained values of Iγ in all cement samples were within the exemption dose criterion (0.3 mSv/y) and corresponds to an activity concentration index of Iγ ≤ 0.5 proposed by  for materials used in bulk construction. The results obtained for Iα in all studied cement samples are lower than the unity. This indicates that the radon exhalation from cement would cause indoor concentration <200 Bq*cm -3 .
The estimated indoor gamma dose rate (D) values for cement samples are shown in [Table 3]. The values obtained in all the studied samples ranged from 47.6 ± 2.2 to 92.4 ± 3.0 nGy/h. These estimated values of indoor gamma dose rate in the studied samples are comparably higher than the world average (populated-weighted) indoor absorbed dose rate of 60 nGy/h recommended by  with the exception of the indoor absorbed dose rate values obtained in CW1 and CW2 cement samples which are relatively lower than the world mean value.
Furthermore, [Table 2] presents also the evaluated AEDE from indoor terrestrial gamma radiation for the studied cement samples. The values obtained varied from 0.2 ± 0.0 to 0.5 ± 0.0 mSv/y. The mean value found to be less that the average external annual effective dose of 0.4 mSv from natural indoor radiation sources to terrestrial.  Comparing the estimated values in the investigated samples with the worldwide average value, it can be observed that the estimated values in Portland, CW4 and CW5 cement sample are slightly higher while those in CW1, CW2 and CW6 are below the recommended values. Comparing the estimated effective dose with the recommended value of 1 mSv/y by International Commission on Radiological Protection (ICRP),  all the values are within the recommended values.
| Conclusions|| |
The specific activities of 226 Ra, 232 Th and 40 K in the studied cement types used in Cameroon were measured using gamma spectrometry based BEGe6350. The observed mean activity values of 232 Th and 40 K are comparably lower than the typical world mean activity of 42 and 420 Bq/kg. However, the mean specific activity values of 226 Ra were lower than the world average value of 32 Bq/kg with the exception of the observed values in CW3, CW4, CW5 and CW6 cement type which were relatively higher than the recommended average value. The observed specific activities of 226 Ra, 232 Th and 40 K were also compared with the reported values by other countries and found to be within the same range. The Ra eq obtained in this study are comparably lower than the world-wide average values of 370 Bq/kg. Absorbed dose and the indoor annual effective dose for an individual living in a house made with the investigated cement are comparatively lower than the recommended average values with the exception of some values obtained in some investigated cement which are comparably higher than the world-wide average value given by UNSCEAR.  From the obtained results, it can be concluded that the investigated cement can safely been used for building construction even though some radiological health hazards parameters are slightly higher than the world-wide reported values.
Since only a few cement type used as building materials have been investigated in this current study, it is suggested that similar investigation should be carried out in other cement type and other building materials used in different parts of the country to have more representative values for the level of naturally occurring radioactive materials, which could be useful to draw a national global picture of radioactivity on building materials in Cameroon.
| Acknowledgments|| |
The authors are grateful for the support and technical cooperation provided by the National Radiation Protection Agency of Cameroon in granting access to the facilities to successfully complete this study.
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[Table 1], [Table 2], [Table 3]