Investigating the Effect of Gamma and Neutron Irradiation on Portland Cement Provided with Waste Silicate Glass

Abstract

In this study, samples of commercial Portland cement mixed with 30% weight of crushed waste silicate glass were prepared in the shape of well-dried cylinders. Then, their physical and mechanical properties were investigated for two types of samples: samples without exposure and samples with exposure to gamma-ray and neutron irradiation. A notable deterioration of the physical properties of the irradiated samples relative to the non-irradiated ones was recorded. All the spectroscopic analyses were performed for the samples with exposure and without exposure to gamma-ray and neutron irradiation. The XRD emerging peaks of irradiated samples were studied to estimate the presence and stabilities of major peaks indicating the presence of the main compositions of cement with the amorphous nature of glass. FT-IR transmittance spectra were identified and the bonds were located close to those of identical glasses. Moreover, SEM images and EDX analysis were conducted on the two types of composite samples (without exposure and with exposure to gamma and neutron irradiation) to specify the change in the physical appearance and the chemical composition after irradiation. The attenuation parameters were computed theoretically with the assistance of Phy-X/PSD software to evaluate the gamma-ray and neutron shielding properties by defining the composition and the density of the samples. The irradiation was found to have a negative impact on the shielding ability of the prepared samples where there was an over-reduction in the parameters calculated with the probability that the damage may increase with longer exposure to the radiation.

1. Introduction

In the past few decades, the technological approaches that employ nuclear radiation have notably increased in many industries, especially the contribution to the field of dosimetry, whether it has been diagnostic, therapeutical, or protectional [1,2,3,4,5,6]. With this growing utilization of nuclear radiation, and because of the nature of the ionizing nuclear radiation which has enough energy to detach electrons from atoms, any leakage is considered as a potential hazard to humans and the surrounding environment. To avoid unwanted damage it is necessary to improve continuously the protection and shielding mechanisms from this type of radiation [7,8,9,10,11].

Selecting a suitable radiation shielding depends on many factors considering the type of radiation, energy, and application [12,13]. Among the most utilized shielding materials are concrete and mortar due to their many advantages including low cost, ease of production, low maintenance requirement, resistance to water, non-toxicity, and most important the malleability to be fabricated in any construction design [14,15,16]. A wide range of various applications have used green additives such as cellulosic waste [17,18], bitumen [19,20], asphaltene [21], glass [22], polymers [23,24,25], nanomaterials [26,27], cement wastes [28,29], and natural clay [30,31] to produce sustainable materials. In spite of the many drawbacks that appear when using concrete and mortar as shielding materials—for instance the density, strength, and cracks that may occur due to long exposure to radiation—concrete and mortar are still widely used as these drawbacks can be limited through the adding of different fillers and admixtures to them. Introducing various compounds into the composition of concrete or mortar may develop a new shield with desirable properties. The most trending admixtures now in the past few years are different waste materials [32,33,34].

Employing waste materials in concrete and mortar has a great benefit to the environment, where these wastes may be hazardous and hard to dispose of. Concrete can work as a vessel of containment for harmful effects due to its inert nature [35,36]. Moreover, it may have a great enhancement in its properties, especially the shielding ones, by adding glass waste to it [37].

Glass is one of the most versatile materials in the world because of its important properties, such as chemical inertness, optical clarity, and high compressive strength. The amount of glass waste generated will increase as the demand for glass components increases.

However, glass waste has become a huge deal because not all glass types can be reused, for example, Pyrex and borosilicate glass, which cannot be considered as recyclable and instead are simply discarded in a careless manner. Although recycling glass in new glass products reduces the use of raw materials and energy consumption, not all glass waste can be recycled in new products because of impurities, mixed colors, or cost. Subsequently, significant studies were conducted in order to use recycled glass as a fine and/or coarse aggregate in mortar and/or concrete because of the good hardness of the glass [14,28]. One of the main routes for glass waste management is adding it into the composition of cement; this can enhance the different properties of the mortar and concrete including the radiation shielding properties as mentioned above. The composite material of waste silicate glass and Portland cement is recommended by the authors [22] in protection from gamma radiation.

The aim of the present work is firstly to prevent the free movement of contaminants in solid glass waste by immobilization of this glass waste in cement as a simple and low-cost technique of containment of the wastes inside a stable solid form in order to be safely and conveniently handled, stored, and disposed. Moreover, secondly, it is to study the possibility of using this composite product as a radiation shielding material and whether if it can withstand excessive irradiation of gamma rays and neutrons or not. Then the aim is to evaluate the damaging impact if present on the mechanical and physical properties of the composite.

2. Materials and Methods

In this experimental study, two distinct materials were used to prepare the intended composite. The first material is ordinary Portland cement (CEMI 42.5 N), which is produced locally by Suez Cement, an Egyptian corporation, and fulfills Egyptian standard ES 4756-1/2005 [38] and British Standard Institution (BSI) EN 197-1/2011 [39]. The second material is waste glass with a high silicate content obtained from broken waste glass (mostly beakers and test tubes). This waste glass was ground into a fine powder for easy mixing with cement.

We used a previously published procedure to prepare the mortar samples [22], with 30% waste glass and 70% cement; 35% water ratio relative to the total weight of the mixture was used to complete the formula. It is worth mentioning that the (30% glass) sample recorded the best mechanical properties [22]. The components were vigorously stirred to achieve a homogenous mortar paste. This paste was shaped into the desired cylindrical form by pouring it into cylindrical plastic molds of the same volume and coating the inside surface with an oily barrier to prevent the mortar from adhering to the mold. The poured mortar mixture was then maintained at room temperature (20° ± 3° Celsius) to allow for a sufficient hydration reaction with water as well as for setting and hardening to complete after 28 days. The water in the samples dried out and the samples solidified. After curing, the solidified samples were taken out from the molds. Different measurements and calculations such as mechanical integrity tests, porosity, spectroscopic analysis, and attenuation were performed twice: before and after irradiation with gamma rays and neutrons.

3. Mechanical Integrity and Porosity Measurements

The mechanical integrity presented in compressive strength was evaluated according to the procedure of the investigated ASTM conditions, and the mechanical performance of solidified specimens was conducted using a hydraulic presser (Ma-Test measuring machine E-159 SP, Italy) for at least three samples [40].

While Archimedes’ saturation method was used to determine the apparent porosity and bulk density, this water displacement technique works based on weighing porous materials both dry and immersed in water [41]. In this experiment there was a uniform dry mass; therefore, specimens were placed in a 105 °C oven for 48 h for complete drying and detecting the dry weight; dry specimens were immersed in boiling water for 2 h to measure the saturated mass; the suspended mass was then estimated by weighing the samples after they had been cured and cooled in water as suspended blocks. The following equation is used to calculate apparent porosity.

$$P(\%)=\frac{W-D}{V}\times 100$$

(1)

where P is apparent porosity, W is the saturated weight of the sample (g), D is the dry weight of the sample (g), and V is the volume of the specimen (cm³) which can be calculated by using the following equation

$$V=W-S$$

(2)

where S is the suspended weight of the specimen (g).

The bulk density ($\rho$), which has the unit g/cm³, of the samples can be determined using its volume and saturated mass as follows:

$$\rho =\frac{D}{V}$$

(3)

4. Irradiation Regime

4.1. The Gamma Irradiation Regime

The samples were subjected to a 200 Gy gamma-irradiation dosage from a Cs-137 source at the Egyptian National Institute of Standards’ Laboratory of Ionizing Radiation Metrology (NIS). The samples were exposed to radiation at a dosage rate of 0.2 Gy/min for about 15 h. The temperature and pressure of the samples was measured using a calibrated thermometer and a parometer before and after irradiation. The absorbed dosage values were computed using an NPL electrometer with an ionization chamber (NE-2561) calibrated at the “Bureau International Des Poids Et Mesures” (BIPM) in France, with a cumulative uncertainty of 0.30%. The dosages were calibrated using air kerma (kinetic energy released per unit mass) at the International Atomic Energy Agency, in accordance with the code of practice guidelines.

4.2. The Neutron Irradiation Regime

The NIS neutron irradiation source was a 5 Ci Am-Be. MCNP5 code simulation was used to calculate the flux and dose rate of the Am-Be source at various distances from the source. At several source-to-detector distances, the dosage rate provided by the source was assessed using an NM2-neutron monitor. Neutron doses were assessed using a PTB-calibrated NM2-neutron monitor. According to the Amersham catalog, the structure of the Am-Be neutron source employed in this work was also simulated using MCNP5 [42,43,44]. The neutron source was a cylindrical capsule with a diameter of 3 cm and a height of 6 cm; the source was encircled by a 1 cm thick poly-methyl methacrylate (PMMA) holder. In this work, the dose received by the samples was 51 millisievert at 1.5 m from the source with a dose rate of 51 millisievert/hour for 1 h.

4.3. Spectroscopic Analysis

The chemical compositions of the mortar samples were determined by using energy-dispersive-X-ray analysis (EDX) in combination with a scanning electron microscope (SEM) unit. A scanning electron microscope was used to investigate the mortar samples to determine the homogeneous distribution of the waste glass in cement.

X-ray diffraction analysis was also used to investigate the mineralogical structure of the samples using a Philips Analytical X-ray system equipped with a Cu-tube anode and a monochromator; measuring was performed at = 1.54056 A°, 2θ range (20°–90°), step size = 0.050°, and step time = 1 s.

Samples were also subjected to a Fourier Transform Infrared spectrophotometer (FT-IR, Tensor 27). Before the spectroscopic investigation, a suitable mass of dried KBr was mixed with ground specimens.

All the spectroscopic analyses were performed for all samples—with exposure and without exposure to gamma and neutron irradiation—after being crushed by a compressive strength test.

4.4. Shielding Parameter Calculation

Materials used mostly for gamma radiation shielding should be consistent in density and composition, as well as thick enough to absorb radiation to a safe level. Gamma-ray attenuation coefficients describe the impact of gamma rays on the matter, with numerous variables which can be used to evaluate the photon attenuation properties of the shielding material, including linear attenuation coefficient, mass attenuation coefficient, half-value layer (HVL), and mean free path (MFP).

The gamma-ray attenuation parameter can be calculated utilizing the famous modified Beer–Lambert law which can be expressed mathematically as follows:

$$I_x=I_o·e^{-\mu x}$$

(4)

where I_x and I_o are the transmitted and the initial photon intensities, respectively, and ($\mu$) is the linear attenuation coefficient (in cm⁻¹), which depends on the incident photon’s energy E, the atomic number (Z), and the material’s density ($\rho$).

The mass attenuation coefficient (${\mu }_m$) can be calculated from the measured values of the linear attenuation coefficient and the density using the following relation:

$${\mu }_m=\frac{\mu }{\rho }$$

(5)

The mean free path (MFP) is the average distance a photon may travel in a medium before interacting with it. The linear attenuation coefficient was used to calculate the mean free path.

$$MFP=\frac{1}{\mu }$$

(6)

The half-value layer (HVL) is the thickness of an interacting material where the intensity of a photon beam entering it is lowered by 50%, and is calculated by the following formula:

$$HVL=\frac{0.693}{\mu }$$

(7)

Radiation protection efficiency (RPE) is an important parameter to consider when evaluating the attenuation qualities of a potential shielding material. Equation (8) may be used to calculate this parameter [45].

$$RPE=\left(1-\frac{I_x}{I_o}\right)\times 100$$

(8)

Because electromagnetic radiation is mainly dissipated by electrons of the atoms, effective atomic number (Z_eff) and effective electron density (N_eff) are another two significant metrics to consider when evaluating gamma radiation shielding. However, Z_eff and N_eff values depend not only on the electron density but also on the photon energy; therefore, the Z_eff and N_eff calculation considers all the electromagnetic interaction processes—photoelectric, Compton interaction, and pair production—that can take place. Evaluating Z_eff and N_eff can be done using the relations in the previous literature [46].

In assessing the Beer–Lambert law in the evaluation of the shielding parameter, this law must obey three important parameters: (1) narrow beam geometry, (2) monochromatic beam, and (3) thin absorption sample. For any deviation from these parameters, that can be in the form of multiple scattering or secondary radiation rather than the primary one, a correction value must be introduced to the law which is called the build-up factor (B). It is known that the value of the build-up factor is equal to or greater than unity. For more accurate measurements of the shielding parameters the build-up factor must come close to unity and the more deviation from the law the greater the value the build-up factor may have. The build-up factor has two forms: the exposure build-up factor (EBF) and the energy absorption build-up factor (EABF). The procedures for calculating EBF and EABF are the same; geometric progression (G-P) fitting is a method for calculating the build-up factors; the process of calculating these parameters is described in the literature [47,48].

The attenuation parameters were computed theoretically with the assistance of Phy-X/PSD which is an online software that is used to evaluate the gamma-ray and neutron shielding properties by defining the composition and density of the samples [49].

The composition values of the samples by weight fraction obtained from the EDX characterization were introduced and the photon energy range of 0.015–15 MeV in addition to selected gamma photon energies obtained from specific radioactive sources provided by the software (²⁴¹Am, ¹⁵²Eu, ¹³⁷Cs, ⁶⁰Co), then all the gamma shielding parameters and fast neutron removal cross-sections (Σ_R) were calculated [50] for samples with exposure and without exposure to gamma and neutron irradiation.

5. Results and Discussion

5.1. Durability and Permeability Tests

The irradiation affected the samples as shown in Figure 1, which could be seen through the increment in the pores and voids percentage in the samples, almost 13% and 12% increase in the case of gamma and neutron irradiation respectively, and the additional pores affected the compressive strength.

The compressive strength for the samples without irradiation exhibited a value of 19.72 MPa; for the samples exposed to irradiation with gamma, the compressive value noticeably decreased to 15.97 MPa; the same behavior occurred in the case of neutron-irradiated samples where the value decreased to 16.36 MPa.

The irradiation also affected the density of the samples and can be seen as a reduction in the values in Figure 2. This reduction can affect the usefulness of the sample in radiation protection.

5.2. SEM Morphological and Microstructure Analysis

SEM/EDX analysis was conducted on samples with exposure and without exposure to gamma and neutron irradiation. The effect of radiation on the material is displayed in Figure 3; the sample without exposure has a smooth surface while the samples subjected to irradiation have developed obvious cracks. This will be discussed and explained in detail in the next two sections.

The chemical composition of the three samples obtained from EDX analysis are shown in

Table 1. Chemical composition of glass/cement mortar samples without exposure and with exposure to gamma and neutron irradiation using EDX.

Element	Glass/Cement Composite Samples			Glass Sample (Without Irradiation)
	Without Irradiation	With Neutron Irradiation	With Gamma Irradiation
C	9.81 ± 0.18	10.10 ± 0.18	11.51 ± 0.19
O	51.38 ±0.66	52.05 ± 0.63	50.67 ± 0.62	58.34 ± 0.36
Na	0.53 ± 0.07	0.68 ± 0.07	0.58 ± 0.06	3.55 ± 0.11
Mg	0.20 ± 0.04	0.42 ± 0.05	0.58 ± 0.05
Al	1.64 ± 0.08	2.13 ± 0.09	1.72 ± 0.08	1.4 ± 0.06
Si	9.37 ± 0.17	8.39 ± 0.16	7.65 ± 0.14	36.17 ± 0.25
S	0.69 ± 0.05	0.91 ± 0.05	1.04 ± 0.05
Ca	24.78 ± 0.29	23.33 ± 0.27	23.95 ± 0.27
Fe	1.59 ± 0.11	2.01 ± 0.12	2.29 ± 0.12
Cu	nd	nd	nd	0.55 ± 0.08
Total	100.00	100.00	100.00

. There was a reduction in the percentage of calcium (Ca) and silicon (Si). In addition, the carbon (C) and sulfur (S) content increased. The change in chemical composition is due to the uneven distribution of the materials of cement and glass with the average values from the EDX.

5.3. The Damaging Impact of Gamma Irradiation

Gamma radiation induces effects through electronic excitations in materials. The excess energy of electrons in excited states is converted to vibrational energies of the constituent atoms of the materials. Consequently, covalent bonds between atoms are broken and radicals are produced. In addition, elevated temperature lingers with long exposure to gamma radiation, which leads to the degradation of the water content and drying out of the mortar samples [51,52].

For the waste silicate glass, gamma irradiation can impact the surrounding and medium-order ranges of the glass network through the accumulation of point defects caused by electronic collisions which may cause displacement cascades, imposing severe stress on the surrounding matrix. Eventually, this can be seen as bond cleavage of some Si-O-Si chains in the network to generate non-bridging oxygens (NBOs), residing in interstitial sites to form new defects, causing the network to become more brittle [53]. The formation of micro-cracks or macroscopic fractures can be seen through the SEM images, which also affect the compressive strength of the specimens [54,55]. This is obviously seen in the present investigation in Figure 3b above.

5.4. The Damaging Impact of Neutron Irradiation

Generally, neutrons have no directly ionizing impact since they are electrically neutral. Neutrons interact directly with the nucleus through scattering interactions (elastic and inelastic scattering) or by absorption, and in this context concrete and mortar are excellent neutron moderators [56] because they are rich with light atoms such as hydrogen, oxygen, and carbon; and when a neutron collides with a lighter nucleus, a neutron is scattered elastically, retaining the nucleus’ overall kinetic energy. Therefore, you don’t see much negative impact on cement from neutron irradiation except for long-term service life [54,57].

However, just like gamma radiation, neutron irradiation can also affect the structure of the waste glass matrix incorporated into the cement, leading to change in the physical and mechanical integrity, and the formation of cracks which can be seen in the SEM images in Figure 3c in agreement with the literature [58].

The overall damage impact is more from gamma irradiation than neutron irradiation, as seen from the compressive strength and porosity (Figure 1), because the neutrons have small or no impact on the mortar unlike the gamma irradiation which damages both the mortar and the glass waste additive.

5.5. XRD Spectroscopic Analysis

The X-ray diffraction examination depicted in Figure 4 shows cementitious material’s major products, portlandite [Ca(OH)₂] and calcite [CaCO₃], at 29° and 34° respectively, as well as other compounds. The three curves are vertically displaced for sake of clarity. In addition, the observed data are consistent with those found in earlier literature for portlandite and calcite [59].

5.6. FT-IR Spectroscopic Analysis

The analysis of FT-IR was conducted to confirm the molecular structure of the samples and to investigate changes between samples without exposure and with exposure to gamma and neutron irradiation. The three curves are vertically displaced for the sake of clarity. Figure 5 proves the presence of portlandite (Ca(OH)₂), which is well identified by distinct peaks at about 3640 cm⁻¹ in all samples, owing to the telescopic stretching vibration of the hydroxyl group. The peaks at 2923.56 cm⁻¹ and 2875.02 cm⁻¹ are associated with CH₂, including CH₂ in the methyl and methine groups, as well as on the aromatic rings. The sharpness of the peaks at 2923.56 cm⁻¹ and 2875.02 cm⁻¹ increased after irradiation with gamma and neutrons which indicates the deterioration of samples and breaking of the cement matrix. Furthermore, cementation materials with peaks between 984.48 cm⁻¹ and 1105.98 cm⁻¹ show the presence of calcium silicate hydrate (C–S–H), which is caused by the Si–O asymmetric stretching vibration. Peaks at 710.64 cm⁻¹, 876.48 cm⁻¹, and 1431.89 cm⁻¹ show the presence of various carbonate group vibration modes. Peaks at 1646.91 cm⁻¹ and 3442.31 cm⁻¹ reflect the vibration of the capillary water’s O–H group. The small peak at 1105.98 cm⁻¹ relates to sulfate, whereas the sharp peak at 467.65 cm⁻¹ reflects silicate in the waste glass [53,60,61].

5.7. Shielding Parameter Results

The mass attenuation coefficient values (computed theoretically) are displayed in Figure 6 at the energy range of 0.015–15 MeV in addition to selected gamma photon energies obtained from specific radioactive sources (²⁴¹Am, ¹⁵²Eu, ¹³⁷Cs, ⁶⁰Co). The computed values show that the mortar sample without irradiation exhibits higher values than samples exposure to gamma and neutron irradiation. At low energies the reduction starts from 2.33% and 1.98% for samples with gamma irradiation and neutron irradiation, respectively, compared to the one without irradiation and this percentage reduced as photon energy increased.

The half value layer percentage shown in Figure 7 increased by 3.73% and 3.85% at 0.662 MeV for samples exposed to neutron irradiation and gamma irradiation, respectively. For the mean free path shown in Figure 8 which represents the average distance between two successive interactions of incident gamma photons, it also shows an increment as the photon energy increased.

RPE% in Figure 9 was calculated for each of the three samples without and with exposure to irradiation and was plotted as a function of photon energy for the three selected energies 0.662, 1.17, and 1.33 MeV. The efficiency reduced by ~3.35% at the three energies selected for the sample with exposure to neutron irradiation and by ~3.43% for the sample with exposure to gamma irradiation in comparison to the sample without irradiation. In addition, for all tested energies, RPE% decreased as energy increased. This decreasing behavior was caused by higher-energy photons easily passing through the sample, reducing the number of photons attenuated by the glass shield and the RPE% as well.

Effective atomic number and effective electron density values as functions of photon energy increase in the range of 0.015–15 MeV are shown in Figure 10 and Figure 11 respectively. Greater values were found in the lowest energy regions, where the photoelectric effect dominates. These values started to decrease as photon energy increased where the curve become nearly flat, in which the Compton effect dominates, and after 1 MeV the values started to increase slightly where the pair production process dominates. The sample without irradiation had higher values for both Z_eff and N_eff than the irradiated ones through all the energy range.

The exposure build-up factor and the energy absorption build-up factor in Figure 12 and Figure 13 show the values calculated at energies of 0.015–15 MeV as functions of different penetration depths of 1, 3, 5, 10, 15, 20, 25, 30, 35, and 40 mfp with the access of the G–P fitting approach.

Table 2. Exposure build-up factor G-P fitting coefficients of mortar samples without exposure and with exposure to gamma and neutron irradiation.

Photon Energy (MeV)		Without Irradiation				With Gamma Irradiation					With Neutron Irradiation
	a	b	c	d	Xk	a	b	c	d	Xk	a	b	c	d	Xk
1.50 × 10−2	0.251	1.024	0.358	−0.171	12.694	0.256	1.024	0.354	−0.175	12.354	0.255	1.024	0.355	−0.174	12.425
2.00 × 10−2	0.180	1.051	0.425	−0.108	17.549	0.179	1.050	0.427	−0.105	17.313	0.179	1.050	0.426	−0.106	17.432
3.00 × 10−2	0.215	1.172	0.395	−0.116	14.068	0.215	1.168	0.395	−0.116	14.116	0.215	1.169	0.395	−0.116	14.099
4.00 × 10−2	0.190	1.362	0.453	−0.107	14.362	0.191	1.354	0.451	−0.107	14.374	0.191	1.357	0.452	−0.107	14.369
5.00 × 10−2	0.146	1.596	0.555	−0.079	14.943	0.149	1.584	0.548	−0.081	14.893	0.148	1.589	0.551	−0.080	14.915
6.00 × 10−2	0.104	1.825	0.669	−0.057	15.045	0.107	1.808	0.661	−0.059	14.985	0.106	1.816	0.664	−0.058	15.013
8.00 × 10−2	0.073	2.288	0.792	−0.051	13.676	0.076	2.264	0.782	−0.051	13.468	0.074	2.276	0.787	−0.051	13.572
1.00 × 10−1	0.024	2.504	0.968	−0.036	13.917	0.026	2.480	0.959	−0.038	13.969	0.025	2.493	0.964	−0.037	13.942
1.50 × 10−1	−0.031	2.643	1.218	−0.012	11.336	−0.029	2.628	1.208	−0.013	11.501	−0.030	2.637	1.214	−0.012	11.409
2.00 × 10−1	−0.050	2.608	1.331	−0.008	8.759	−0.049	2.597	1.323	−0.009	8.929	−0.050	2.603	1.328	−0.008	8.831
3.00 × 10−1	−0.073	2.438	1.438	0.013	18.849	−0.072	2.431	1.433	0.013	19.152	−0.073	2.436	1.436	0.013	18.969
4.00 × 10−1	−0.076	2.328	1.443	0.014	17.031	−0.075	2.324	1.439	0.014	17.261	−0.075	2.327	1.442	0.014	17.100
5.00 × 10−1	−0.076	2.226	1.435	0.016	16.309	−0.076	2.223	1.432	0.016	16.309	−0.076	2.225	1.434	0.016	16.310
6.00 × 10−1	−0.075	2.149	1.417	0.018	16.595	−0.075	2.146	1.415	0.018	16.535	−0.075	2.148	1.416	0.018	16.483
8.00 × 10−1	−0.071	2.036	1.376	0.021	15.618	−0.071	2.035	1.374	0.020	15.703	−0.071	2.035	1.375	0.021	15.636
1.00	−0.063	1.960	1.323	0.020	16.948	−0.063	1.958	1.322	0.020	16.941	−0.063	1.959	1.323	0.020	17.008
1.50	−0.048	1.860	1.231	0.017	15.403	−0.048	1.859	1.231	0.017	15.524	−0.048	1.859	1.231	0.017	15.556
2.00	−0.033	1.788	1.155	0.011	15.981	−0.033	1.787	1.155	0.011	16.072	−0.033	1.787	1.155	0.011	16.083
3.00	−0.012	1.677	1.062	−0.001	14.955	−0.013	1.676	1.062	−0.002	15.236	−0.013	1.675	1.063	−0.002	15.614
4.00	0.005	1.600	0.995	−0.009	12.970	0.005	1.600	0.996	−0.009	12.977	0.005	1.600	0.996	−0.009	12.988
5.00	0.021	1.537	0.946	−0.021	11.818	0.022	1.537	0.945	−0.021	11.594	0.022	1.538	0.944	−0.021	11.260
6.00	0.026	1.487	0.927	−0.021	11.795	0.026	1.487	0.928	−0.021	11.822	0.026	1.486	0.929	−0.020	11.864
8.00	0.034	1.405	0.902	−0.028	13.736	0.034	1.405	0.902	−0.028	13.750	0.034	1.405	0.902	−0.028	13.772
1.00 × 10+1	0.042	1.344	0.882	−0.033	13.126	0.042	1.344	0.883	−0.033	13.122	0.042	1.343	0.883	−0.033	13.115
1.50 × 10+1	0.060	1.257	0.838	−0.053	14.281	0.059	1.256	0.839	−0.053	14.277	0.059	1.256	0.841	−0.052	14.269

and

Table 3. Energy absorption build-up factor G-P fitting coefficients of mortar samples without exposure and with exposure to gamma and neutron irradiation.

Photon Energy (MeV)		Without Irradiation				With Gamma Irradiation					With Neutron Irradiation
	a	B	c	d	Xk	a	b	c	d	Xk	a	b	c	d	Xk
1.50 × 10−2	0.200	1.023	0.405	−0.096	11.759	0.195	1.023	0.410	−0.087	11.474	0.196	1.023	0.409	−0.089	11.533
2.00 × 10−2	0.180	1.051	0.424	−0.108	17.559	0.180	1.050	0.425	−0.107	17.292	0.180	1.050	0.425	−0.107	17.418
3.00 × 10−2	0.216	1.173	0.394	−0.118	14.302	0.216	1.168	0.394	−0.119	14.241	0.216	1.170	0.394	−0.119	14.263
4.00 × 10−2	0.191	1.375	0.450	−0.106	14.626	0.191	1.366	0.448	−0.106	14.632	0.191	1.370	0.449	−0.106	14.629
5.00 × 10−2	0.134	1.634	0.572	−0.070	16.189	0.139	1.622	0.562	−0.073	16.001	0.137	1.627	0.566	−0.071	16.083
6.00 × 10−2	0.153	2.085	0.571	−0.082	13.627	0.156	2.062	0.563	−0.084	13.644	0.155	2.073	0.566	−0.083	13.636
8.00 × 10−2	0.082	2.892	0.766	−0.058	13.721	0.086	2.857	0.755	−0.060	13.664	0.084	2.874	0.761	−0.059	13.693
1.00 × 10−1	0.026	3.547	0.962	−0.036	13.719	0.029	3.515	0.949	−0.037	13.709	0.028	3.531	0.955	−0.036	13.714
1.50 × 10−1	−0.047	3.974	1.280	0.003	18.386	−0.044	3.962	1.268	0.001	19.160	−0.046	3.969	1.275	0.002	18.728
2.00 × 10−1	−0.073	3.704	1.430	0.015	15.796	−0.070	3.708	1.417	0.013	16.031	−0.072	3.706	1.425	0.014	15.894
3.00 × 10−1	−0.092	3.129	1.536	0.024	14.201	−0.091	3.132	1.528	0.023	14.162	−0.092	3.130	1.533	0.024	14.186
4.00 × 10−1	−0.093	2.803	1.532	0.026	15.162	−0.092	2.806	1.526	0.025	15.277	−0.092	2.804	1.529	0.025	15.196
5.00 × 10−1	−0.091	2.588	1.507	0.026	15.265	−0.090	2.590	1.503	0.025	15.309	−0.090	2.589	1.505	0.025	15.287
6.00 × 10−1	−0.087	2.442	1.477	0.026	14.945	−0.087	2.443	1.474	0.026	14.980	−0.087	2.443	1.475	0.026	14.941
8.00 × 10−1	−0.078	2.245	1.411	0.024	15.202	−0.078	2.245	1.409	0.024	15.221	−0.078	2.245	1.410	0.024	15.209
1.00	−0.071	2.116	1.358	0.024	15.008	−0.071	2.116	1.357	0.024	15.051	−0.071	2.115	1.358	0.024	15.010
1.50	−0.052	1.940	1.247	0.020	14.634	−0.052	1.940	1.246	0.020	14.652	−0.052	1.940	1.246	0.020	14.656
2.00	−0.036	1.835	1.164	0.013	14.445	−0.036	1.834	1.165	0.013	14.428	−0.036	1.834	1.165	0.013	14.436
3.00	−0.011	1.699	1.058	0.000	13.729	−0.011	1.698	1.059	0.000	13.924	−0.012	1.697	1.059	0.000	14.185
4.00	0.007	1.609	0.989	−0.010	14.056	0.007	1.609	0.990	−0.010	14.114	0.007	1.608	0.990	−0.010	14.198
5.00	0.026	1.547	0.930	−0.023	12.905	0.026	1.547	0.929	−0.024	12.921	0.027	1.547	0.927	−0.025	12.944
6.00	0.025	1.476	0.929	−0.028	15.608	0.025	1.475	0.929	−0.028	15.584	0.025	1.475	0.929	−0.028	15.547
8.00	0.032	1.383	0.907	−0.023	12.074	0.032	1.382	0.907	−0.023	12.060	0.032	1.382	0.908	−0.023	12.038
1.00 × 10+1	0.032	1.314	0.909	−0.026	14.337	0.031	1.314	0.910	−0.026	14.364	0.031	1.313	0.912	−0.026	14.408
1.50 × 10+1	0.050	1.229	0.862	−0.045	14.370	0.051	1.229	0.860	−0.045	14.346	0.053	1.229	0.857	−0.047	14.308

show the G-P parameters for the exposure build-up factor and the energy absorption build-up factor, respectively.

For each of the three samples without and with irradiation, at low energies EBF and EABF values are low, when photoelectric effect is dominant and elimination of almost all the incident gamma photons occur. As the energy rises the Compton process starts to dominate and only a small fraction of the incident photon energy is removed. Scattered photons and the remainder of the incident photon energy accumulated within the material and the values of EBF and EABF rise in the mid-energy range. At high energy above 1 MeV, which is the threshold of the pair production process, the values of EBF and EABF start to decrease gradually.

The calculated EBF and EABF values were found to increase sequentially as the penetration depth increased up to 40 mfp. Gamma photons take much longer time to penetrate the thicker layer, accumulating more photons. The lowest EBF and EABF values are obtained at low penetration depth (1 mfp), while the highest EABF values are obtained at high penetration depth (40 mfp). In addition, EBF and EABF for the sample without irradiation have higher values than for the samples irradiated with gamma and neutrons.

It is well understood that neutrons have a greater biological effect than photons, necessitating the use of suitable attenuators for safety. Figure 14 shows the values of the fast neutron removal cross-section without and with irradiation, in which there is no great effect of irradiation on the tested samples in agreement with literature [62].

6. Conclusions

In this work, the physical properties for testing the shielding capacity of prepared composites of cement and waste glass were measured for three samples: one without exposure, and two with exposure to gamma-ray and neutron irradiation. The results demonstrate that there is a negative impact of irradiation on the samples through the reduction of the integrity of the samples as shown in the compressive strength test.

Sample densities after irradiation show a reduction that can affect the usefulness of the samples in radiation protection.

SEM/EDX analysis was conducted on the samples (without exposure and with exposure to gamma and neutron irradiation) too. The figures show obviously the effect of radiation, as the sample without irradiation exposure has a smooth surface while the samples subjected to irradiation have developed obvious cracks.

The irradiation also has a negative impact on the shielding ability of the prepared samples and there is an over-reduction in the parameters calculated with the probability that damage may increase with longer exposure to the radiation.