Introduction

The application of radiation sources in various fields and the use of gamma ray-emitting isotopes in nuclear research, industry, medicine, agriculture and education in the world was increasing1. When radiation interacts with material, it loses some of its energy by deposition into the atoms of the matter. Ionizing radiation such as gamma-radiation sources that are used in nuclear medicine, contribute to improve people’s quality of life, it also represents a risk2. Ionizing radiation causes the ionization of the cell's atoms, thereby destroying the normal chemical equilibrium of the cell and eventually can lead to the cell's death or permanent deformation3.

Terrestrial radionuclides that can be found on Earth typically emerged during the formation of the planet4,5 and some of them takes several years to decay and cease to be radioactive, but they remain a part of the system4,5. They are known to originate in the earth and enter the environment through soil, air, water, and building materials5,6. It stands to reason those natural radionuclides such as 238U, 232Th and 40K are contains in these materials at different proportions, and are used for different purposes radiation shielding inclusive5,7,8.

Since radiation can pose a serious concern in medical X-ray systems, industries, nuclear reactors and many other instances, then radiation shielding is imperative, in order to protect the general public and workers involved in radiation related fields from the radiation effects and patients exposed to ionizing radiation for treatment purposes9. Containing radiation and preventing it from causing physical harm to patients, employees, and or their environment is an integral part of operating equipment that emits radiation. Preserving human and structural material that may be compromised from radiation exposure is very necessary with different radiation-shielding materials better suited than others as determined by the interaction between radiation and the elemental properties of the shielding materials5,10.

Radiation shielding is based on the principle of attenuation, which is the ability to reduce the dose by completely absorbed or scattered when passed through a material medium. When radiation interacts with a material it loses some of its energy through different kinds of interactions such as photo-electric effect, Compton’s effect and pair production3. Radiation shielding, which involves using materials to reduce radiation intensity and protect against radiation damage, is essential in various fields. For decades, lead has been considered as the most often use in radiation shielding. It is commonly used for X-rays and gamma radiation shields because of its high density, high atomic number and therefore with high linear attenuation coefficients11. It is easy to process and provides durable shielding, but yet there has been an overwhelming increase in health, safety, and environmental concerns over the mining, processing, handling, and its disposal5. Lead can be hazardous when taken into the body by swallowing or breathing in lead or materials contaminated with lead. In fact, lead has already been banned from use in many applications, such as motor fuels, paint, and water pipes in many countries and its total prohibition in several countries12.

Recognizing the need for an alternative shielding material, researchers have sought for materials that are nontoxic, environmentally friendly, and capable of providing reliable radiation protection13,14,15. For decades, local building materials, such as clay, gypsum, kaolin, and sand were used as raw materials for building in the world, research have shown that these materials can be used for the purpose of radiation shielding in order to protect the general public and the personnel from the harmful effects of radiation. They were found to be promising candidates and offer potential substitutes for lead due to their high density. These materials are abundant in various quantities across the study area. Therefore, there is always a need to develop new and improved materials that can be used under potential harsh conditions of radiation exposure and act as shielding materials for extended periods of time15,16,17.

For the selection of an appropriate material for gamma ray shielding, shield-designers have to investigate materials’ suitability by using its Gamma-ray Shielding Parameters (GSP) (GSP)10,18. Garba et al., in 2023, assessed the radiological status of some commonly used building materials (clay, sand, kaolin and gypsum) in Northwestern Nigeria and suggested that they can further be explored for radiation shielding application5.

Therefore, this study investigates the gamma radiation shielding behaviours of clay, sand, kaolin and gypsum commonly used as building materials in Northwestern Nigeria. For this reason, several parameters such as attenuation effectiveness, HVL, TVL and Zeff were determined.

Materials and methods

Four different locally used building materials (clay, sand, kaolin, and gypsum) from the areas of (Zamfara, Sokoto, Kebbi, Katsina, Kano, Jigawa and Kaduna) in Northwestern Nigeria were obtained with the aid of local suppliers. Then prepared for analysis to ascertain their experimental and theoretical gamma ray shielding behaviours using standard point sources of Cesium-137 and Cobalt-60 of photon peaks (0.662 and 1.133, 1.173) MeV and Py-MLBUF software. During the measurements, thallium activated sodium iodide NaI(TI) detector was used. The elemental compositions of the samples were determined using Neutron Activation Analysis (NAA) at the Centre for Energy Research and Training (CERT) Ahmadu Bello University, Nigeria. The choice of NAA technique was because it is fast, accurate and non-destructive technique.

Sample collection and preparation

The selected materials were collected in plastic polyethylene bags, weighed, packed, and transported to the laboratories at the Physics Department and Centre for Energy Research and Training, Ahmadu Bello University, Zaria for preparation and analysis. The collected samples were oven dried at 105 °C for 24 h in order to remove moisture contents and then allowed to cool down to room temperature and later crushed, grounded and then sieved for homogeneity using 500 μm and 250 μm mesh5,19.

Methodology

Measurement of Activity Concentrations of 238U, 232Th, and 40K

The prepared samples were analysed for the activity concentrations of 238U, 232Th, and 40K using NaI (Tl) gamma ray detector situated at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria. The detector consists of a 7.62 × 7.62 cm NaI (TI) detector housed in a 10 cm thick lead-shield, cadmium-lined assembly with copper sheets for the reduction of background radiation, which was equipped with a MAESTRO computer system program for data acquisition and spectra analysis. IAEA standard reference materials (RGK-1, IAEA-448, and RGTh-1) were used for the quantitative analysis of the 238U, 232Th, and 40K, respectively5,20. Each of the prepared samples was counted for approximately 4 h, with the activity concentration of 238U determined using the gamma energy line of 214Bi (1760 keV), 232Th determined using 208Tl (2614 keV), while the activity concentration of 40K was determined directly from its 1460 keV gamma line. Energy and efficiency calibrations were carried out using a point source of 500 mL of 60Co (1173 and 1332 keV), 241Am (59.54 keV), and 137Cs (661.62 keV) multi-nuclide standard solution, respectively. The net number of counts under each photo peak of interest was then background subtracted using the time-correct spectrum taken using the blank container and the specific activity concentrations of the radionuclides were calculated using Eq. (1)5,21.

$${\text{A}}=\frac{{\text{N}}-{{\text{N}}}_{{\text{o}}}}{{{\text{I}}}_{\upgamma }\mathrm{\varepsilon mt}},$$
(1)

where A is the specific activity of the radionuclide in the sample, N is the net counts of a given peak for a sample, No is the background count of the given peak, Iγ is the number of gamma photon per disintegrations, ε is the detector efficiency at the specific gamma-ray energy, m is the measured mass of the sample, and t is the measuring time5

Measurement of elemental compositions using neutron activation analysis (NAA)

The samples analysed using Nigerian Research Reactor-1 (NIRR-1) at the Centre for Energy Research and Training, Ahmadu Bello University. NIRR-1 is a low-power nuclear reactor, which has highly enriched uranium as fuel, light water as moderator and beryllium as reflector22. The prepared samples were exposed to neutrons causing a portion of the atoms to undergo neutron capture, which produces high-energy compound nuclei that rapidly transform into radioactive forms of the original chemical element(s). As the radioactive isotopes undergo decay to reach stable ground state configurations, the sample is placed on a high-purity germanium detector, which records the intensities and energies of the gamma rays that are emitted23,24. The radioactive emissions and their decay paths for each element are well known. The radioactive isotope always emits gamma rays at certain specific energies and intensities. The radioisotopes present, and hence the parent chemical element(s) present in the sample, was determined quantitatively specified by comparison with the activities of the standard reference materials23,25.

Measurement of gamma ray shielding parameters (GSP)

Information about various gamma-ray shielding parameters (GSP) is required to investigate the gamma-ray shielding efficacy of the materials. The effectiveness of a shield depends on the energy of the photons and on the type of the shielding material. The higher the atomic number and the density of the shielding material the more effective it is in reducing the intensity of gamma radiations.

Theoretical assessment of GSP

For the simulation of gamma ray shielding parameters of the studied samples, Python-program for Multi-Layered Build-Up Factors (Py-MLBUF) computer code was used in computing the gamma ray shielding parameters in the photon energy range of (0.1–5.0) MeV. In achieving these, the samples elemental compositions (by weight) and densities were used as input parameters to the computer code18.

Experimental assessment of GSP

The samples were prepared by mixed with distilled water, and then a homogeneous mixture of each sample was obtained and cast into blocks of 1, 2, and 3 cm thicknesses for each of the samples. After 24 h, the blocks were removed from the moulds and placed in a plastic container. Special care was taken in order to preserve their physical intactness when carrying them to the laboratory for the experiment.

Furthermore, the mass density of the prepared samples was determined from the calculated mass of each sample using a digital beam balance, and the volume of each sample was calculated from the dimensions of the moulded blocks and the density was obtained using Eq. (2).

$$ \rho = \frac{m}{V}\left( {{\text{kg}}/{\text{m}}^{{3}} } \right), $$
(2)

where, m and V are the mass and volume of the sample respectively.

The gamma ray shielding measurements of the prepared samples were performed using NaI (TI) detector of dimension 7.62 × 7.62 cm housed in a 10 cm thick lead-shield at Centre for Energy Research and Training, Ahmadu Bello University Zaria, Nigeria. Cesium-137 and Cobalt-60 gamma ray point sources of photon peaks (0.662 and 1.133, 1.173) MeV were used. The incident and transmitted gamma-ray intensities of the point sources due to the presence of the samples were determined for the measurements of mass attenuation coefficients. A narrow beam transmission geometry condition was adopted for the measurements to ensure minimal scattered radiation as shown in Fig. 1.

Figure 1
figure 1

Experimental set-up of the narrow beam transmission geometry26.

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The total mass attenuation coefficient (MAC), linear attenuation coefficient (LAC), half-value layer (HVL), tenth value layer (TVL) and effective atomic number (Zeff) for the gamma ray shielding of the samples were evaluated using Eqs. (3–6) respectively.

$${\text{MAC}}=\sum {{\text{w}}}_{{\text{i}}}{\left({\text{MAC}}\right)}_{{\text{i}}},$$
(3)

where, \({w}_{i}\) is the weight fraction, (MAC) represent the mass attenuation coefficient of the ith constituent18. The linear attenuation coefficients (LACs) which indicates the rate at which photons interact as they move through material and is inversely related to the average distance photons travel before interacting are calculated Eq. (4)27.

$${\text{LAC}}= \frac{-{\text{In}}(\frac{{\text{I}}}{{{\text{I}}}_{{\text{o}}}})}{{\text{x}}}. $$
(4)

The half-value layer (HVL) of the absorbing material which is a measure of the thickness that decrease the incident photons intensity to half of its initial value is calculated using Eq. (5)27,28.

$$\mathrm{HVL }= \frac{{\text{ln}}2}{{\text{LAC}}}.$$
(5)

The tenth-value layer of the absorbing material which represents the thickness that decrease the incident photons intensity to one tenth of its initial value was calculated using Eq. (6)28,29.

$$\mathrm{TVL }= \frac{{\text{ln}}10}{{\text{LAC}}}.$$
(6)

Additionally, the effective atomic number (Zeff) which is an important parameter in radiation shielding that describes the attenuation of gamma and X-ray photons radiations for any material with high values of it indicating better attenuation of gamma and X-ray photons was calculated using Eq. (7)30.

$$ Z_{eff} = \frac{{\sigma_{a} }}{{\sigma_{e} }}, $$
(7)

where σa and σe are the total atomic and electric cross-section of the materials18,29.

Results and discussion

Activity concentrations of 226Ra, 232Th and 40K and elemental composition

The activity concentrations of 238U, 232Th and 40K, the elemental compositions and densities of the samples determined were presented in Tables 1 and 2 respectively. Generally, it can be observed from Table 1 that, the activity concentration of 40K was found to be the highest in all the analysed samples of the four commonly used building materials within the study area, then followed by 2226Ra with 232Th having the lowest. This clearly indicates that 226Ra and 40K are the major sources of gamma radiation in the studied samples and this is in good agreement with a similar studies conducted in Kerala, India31. Additionally, the relatively high amount of 40K observed in all the samples though not above the world average value, can be attributed to the amount of potassium silicate in the geological formations of area and the intense agricultural activities in the environment5. Therefore, it can be concluded that the samples are radiologically free and thus have the potential to be utilized for radiation shielding purposes.

Table 1 Activity concentrations of 238U, 232Th and 40K.
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Table 2 Elemental compositions and density of the samples.
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From Table 2, it can be seen that high mean concentrations of Si and Al were found to be in kaolin with respective values of 55.03% and 22.25%, while that of K and Fe were found to be in clay with respective values of 12.91% and 19.09%, and that of Ba with an average value of 1.44% was found to be in Sand. High mean concentration of Ca was found to be in gypsum with a value of 96.46% which is in line with the fact that gypsum is known to be rich in calcium and informs its utilization as an essential component of cement32.

Similarly, silicon content was almost the same in all of the kaolin, sand, and clay samples, regardless of where they came from. This is not strange as it is well known that silicon, which is essentially abundant on the outermost layer of the earth's crust, is found in significant quantities in clay, sand, and kaolin32. On the other hand, elements like Mg, K, Ti, Mn, Zn, and Na, varied greatly in concentrations depending on the samples.

Gamma ray shielding parameters results (GSP)

Gamma photon attenuation

Figures 2 and 3 presents the attenuation of photon beams from gamma radiation sources of 137Cs and 60Co peaks at various sample thicknesses. The results indicates that all the samples showed a rising behaviour as the thickness increased from 1 to 3 cm. The materials (gypsum, kaolin, and clay) were observed to respectively attenuates 63.5%, 61.5%, 58.4%, and 44.2% of gamma-ray photons at a thickness of 3 cm for a gamma energy of 0.662 MeV. This was also the case for gamma energy of 1.332 MeV, in which photons were attenuated by 40.6%, 32.9%, 30.6%, and 27.3% respectively. However, gypsum was observed to be the most effective material for significantly attenuating the photon beams followed by kaolin and clay samples. In addition to having a greater concentration of Ca and Fe than any other sample, gypsum was found to have a higher density than the other materials, clay and kaolin, respectively, which contributed to its superior performance. The general behaviour of all samples demonstrates a linear relationship between the sample thickness and the measured photon beam attenuation.

Figure 2
figure 2

Attenuation of gamma rays against the 137Cs energy source as a function of sample thickness.

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Figure 3
figure 3

Attenuation of gamma rays against the 60Co energy source as a function of sample thickness.

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Mass attenuation coefficient (MAC) and linear attenuation coefficient (LAC)

The Mass Attenuation Coefficient (MAC) and Linear Attenuation Coefficients (LAC), were experimentally determined and compared with the theoretically computed results obtained using PY-MLBUF computer code of gamma energy ranges of (0.1–5.0) MeV18 and presented in Figs. 4 and 5, and Tables 3 and 4, respectively as a function of photon energy. It was observed that MAC and LAC decreases as photon energy increases. Similarly, as the photon energy increases from 0.1 MeV to about 0.8 MeV, MAC and LAC approaches comparable values for all the samples. Furthermore, at very low energies, MAC and LAC were observed to sharply drop as energy increases, which was attributed to the high inverse correlation between the photoelectric effect and photon energy in agreement with findings of Singh et al.33.

Figure 4
figure 4

Mass attenuation coefficient as a function of photons energy of the samples.

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Figure 5
figure 5

Linear attenuation coefficient as a function of photons energy of the samples.

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Table 3 Experimental and theoretical mass attenuation coefficient of the samples.
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Table 4 Experimental and theoretical linear attenuation coefficient of the samples.
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It was further observed from Figs. 4 and 5 that the gypsum MAC and LAC values are higher than that of all the samples analyzed in the energy range, this is because MAC and LAC values are influenced by photon energy, density and the atomic number of the materials that are capable of changing photon attenuation in the samples in line with the work of Simon et al.10. However, the high values of MAC and LAC in gypsum could be due to the high content of Fe and Ca, and the high density of the sample compared to others.

Furthermore, Tables 3 and 4 presents a comparison of MAC and LAC results determined experimentally and theoretically using Py-MLBUF.

It can be observed from the tables that the results of the two methods are in good agreements with each other.

Half-value layer (HVL) and tenth value layer (TVL)

The HVL and TVL as a function of photon energy are presented in Figs. 6 and 7.

Figure 6
figure 6

Half-value layer as a function of photons energy of the samples.

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Figure 7
figure 7

Tenth-value layer as a function of photons energy of the sample.

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The figures showed that the HVL and TVL results increases with increment in incident photon energy. It can be observed that at very low energy region (< 0.8 MeV), the HVL and TVL values were almost the same for all the samples, due to the similar chemical contents of all the materials which agrees with the findings of Al-Jaffs34. It was also observed that the higher values of HVL and TVL occur as the energy increase reaches its maximum value at the highest energy region. Furthermore, it could be seen from the figures that as energy increases, the HVL and TVL results approach identical values for all the samples at gamma energy region of (0.1–0.8) MeV. However, as the photon energy increases above 2.0 MeV, a little increment in HVL and TVL values was observed and this behaviour could be due to the high Z-element dependency and the dominance of pair production at higher energy region as earlier stated by Al-Buriahi et al.30. It was also observed that all the samples have virtually same value at a specified gamma energy and this can be attributed to the same chemical compositions of the elements with high atomic number in the samples. Materials with low HVL and TVL were found to give better radiation shielding effectiveness due to low thickness of these materials which agreed with the work of Salisu et al.35. It was observed that HVL and TVL values are influenced by chemical contents, photon energy and density of the materials36. The HVL and TVL results for gypsum are low in comparison with other materials and this notified the highest density property in gypsum over other materials used in this work.

Tables 5 and 6 presents a comparison of HVL and TVL values determined experimentally and theoretically, in which a good agreement in the results of the two methods was observed.

Table 5 Experimental and theoretical half-value layer of the samples.
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Table 6 Experimental and theoretical tenth-value layer of the samples.
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Effective atomic number (Zeff)

Figure 8 present a plot of the effective atomic number Zeff of the four samples against photon energy obtained using the Py-MLBUF program. It was observed that the values of Zeff for all samples were higher at the lower energy region, this is primarily due to the photoelectric interaction which directly depends on the atomic number, Z4, and the photon energy as, E−3.5, for any absorber37.

Figure 8
figure 8

Effective atomic number (Zeff) as a function of photons energy of the samples.

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At the intermediate photon energy, the Zeff decreases slowly with the increase of incident photon energy and becomes almost independent of the photon energy for all the samples which agrees with the work of Akman et al.38. As the photon energy increases above 3.0 MeV, the value of Zeff showed a little increment and this behaviour could be due to the high Z-element dependency and the dominance of pair production at higher energy region. At low energy region, the highest value of Zeff was observed for gypsum and the minimum value was observed for clay sample, this indicated that Zeff is linearly related to the Z-elements in the sample as earlier reported by Akman et al.38. Generally, it was observed that the Zeff values of the samples decreased with increasing incident photon energies.

Conclusion

Gamma-ray shielding parameters of locally used building materials (clay, sand, kaolin, and gypsum) in the Northwestern Nigeria were determined experimentally using 60Co and 137Cs gamma radiation standard point sources of photon peaks (0.662 and 1.133, 1.173) MeV and theoretically using Py-MLBUF software. The results showed that MAC and LAC values of the samples decrease with an increase in photon energy while HVL and TVL values increases. The experimental data of all the parameters were found to be in good agreement with the theoretical results obtained using Py-MLBUF. The studied samples (gypsum, kaolin, sand, and clay) were found to attenuate 63.5%, 61.5%, 58.4%, and 44.2% of gamma-ray photons with an energy of 0.662 MeV at a thickness of 3 cm, and 40.6%, 32.9%, 30.6%, and 27.3% of gamma energy of 1.332 MeV respectively. However, gypsum was found to have the best shielding behaviours and this is attributed to its high density (2.28 g cm−3) determined compared with the rest of the samples. The locally used building materials from the Northwestern Nigeria particularly gypsum can be used in shielding design and constructions of storage facility for X-ray and gamma ray sources and other low energies radioactive materials.