Diagnostic and therapeutic radioisotopes in nuclear medicine: Determination of gamma-ray transmission factors and safety competencies of high-dense and transparent glassy shields

1 Introduction

Nuclear medicine is a field of research in which radioisotope-containing medications are used to diagnose and treat diseases [1]. It began in the 1950s with the use of 131-Iodine for thyroid cancer detection and treatment. Nuclear medicine techniques are extremely effective, reliable, and painless. They involve administering a small amount of radioisotope material or radiopharmaceuticals (inorganic compounds labeled with radioisotopes, organic compounds, peptides, proteins, monoclonal antibodies and fragments, and oligonucleotides) to the patient to examine the physiological and molecular processes occurring in the body. On the other hand, gamma-rays generated by radioisotopes administered to patients are detected in nuclear medicine using planar or tomographic technologies, such as gamma camera, single photon emission tomography (SPECT), positron emission tomography (PET), and hybrid systems (SPECT/CT, PET/CT). Diseases are diagnosed using the images generated as a consequence of processing [2,3]. In nuclear imaging, radioisotope-labeled carriers are used as noninvasive diagnostic tools to offer information on the function and structure of tissues and their surroundings [4]. All radioisotopes used in nuclear medicine are synthetic, and they are manufactured in a variety of ways, including fission in the reactor, thermal neutrons in the reactor, generator-made radioisotopes, and cyclotron-produced radioisotopes [1]. Radioisotopes are utilized in nuclear medicine for two purposes, such as therapy and diagnostics. While radioisotopes that generate electromagnetic gamma radiation are utilized for imaging, radioisotopes that are heavier, have a larger ionization energy, and decay by scattering beta- or alpha-rays with a particulate nature are employed for therapy. Meanwhile, nuclear medicine is an intentional exposure to radiation. While ionizing radiation travels through living tissue, it transfers part or all of its energy to the tissue, resulting in recognized detrimental effects on live creatures at low to high exposure levels. These are stochastic (cancer, mutations) and deterministic (tissue responses) consequences (such as dermatitis, cataract). It is critical to maintain radiation workers and the general public’s exposure to radiation below safe dosage limits to avoid undiscovered negative consequences. ALARA (as low a dose as reasonably attainable) and ALARP (as low a dose as reasonably practical) principles, as well as dosage limitations established for the profession and society by international and national atomic energy institutes, and radiation safety legislation, should be obeyed [3]. With the introduction of novel targeted radioisotopes for treatment, particularly for malignant disorders, radioisotope therapy is increasingly used in nuclear medicine clinics [4]. As a result, while building nuclear medicine units, certain objectives should be defined based on time, distance, and shielding parameters to minimize radiation exposure to both personnel and patients. These objectives include assuring the safety of radioactive sources, maximizing employee, patient, and public exposure, maintaining complete control over radioisotope/radiopharmaceutical activities, and avoiding contamination spread [5,6,7]. Therefore, the scientific community has concentrated on developing next-generation shielding materials that may provide certain advantages over conventional shielding materials in terms of avoiding those shortcomings. In this study, we presented the findings of a complete examination of newly designed CdO-rich and transparent glass shields for usage in nuclear medicine facilities in lieu of traditional and unfavorable materials, such as lead (Pb) and concrete. Gamma-ray transmission factors (TFs) of newly designed glass shields are determined using a variety of diagnostic, therapeutic, and research radioisotopes, including ⁶⁷Ga, ⁵⁷Co, ¹¹¹In, ²⁰¹Tl, ^99mTc, ⁵¹Cr, ¹³¹I, ⁵⁸Co, ¹³⁷Cs, ¹³³Ba, and ⁶⁰Co. The study’s findings may provide key information for optimizing shielding materials used to protect personnel and patients in units, such as nuclear medicine, that utilize medical radiation sources.

2 Materials and methods

2.1 Investigated glassy shields

Cd-doped glasses were effectively fabricated using the melt-quenching process and a range of CdO compositions (x = 0, 15, 20, 30, and 40 mol%) (see Table 1). Each oxidized chemical was separately weighed using a high-precision scale. The platinum crucible containing the solution was heated to very high temperatures during the synthesis of glass shields. The high-temperature furnace was permitted to warm up to 900°C from room temperature for the first 60 min of the procedure. Mechanical stirring was performed every 15 min throughout this time period. Each glass sample was annealed for about an hour at 380–385°C and then cooled back to room temperature (Figure 1).

Table 1

Glass codes, glass compositions, thickness, and density values of glasses

Code	P2O5 (mol%)	TeO2 (mol%)	ZnO (mol%)	CdO (mol%)	Density (g/cm3)	Thickness (mm)
C0	20	30	50	—	4.41970	2.637
C15	20	30	35	15	4.65192	2.678
C20	20	30	30	20	4.72403	2.66
C30	20	30	20	30	4.87138	2.785
C40	20	30	10	40	5.01752	2.447

Figure 1

Synthesized glass series.

2.2 Calculation of gamma-ray TFs

The absorption characteristics of materials intended for use in fields containing medical radiation, such as nuclear medicine and diagnostic radiology, against gamma-rays are significant. The reason for this is the exposure that personnel working in such environments may face when dealing with the associated radiation sources, as well as the deterministic and stochastic impacts, which this exposure may have in the short and long terms [8,9,10,11,12,13,14]. The TF [15,16,17] is a critical shielding metric that is derived using the percent reduction of primary gamma-ray radiation incident on a material. It demonstrates the length dependency of the gamma-ray transmission process. Calculating TF values for a single material offers information on the degree of absorption provided by that material, while studying it for a group of materials provides critical information about the effect of changing material content on this transition factor. Thus, the direct influence of manufacturing modifications on the absorption characteristics is also well recognized. In this study, the absorption properties of CdO-based and high-density glass materials produced against some diagnostic and therapeutic radioisotopes (see Table 2) were investigated for different energies and different material thicknesses and compared with some conventional shielding materials within the framework of the same parameters. The general-purpose radiation transport code MCNPX [18] operating with the Monte Carlo simulation method was used to calculate the TF values. First, the glass structures produced were defined according to the elemental properties given in Table 1 in the INPUT file of the MCNPX code, and their respective densities were included in the INPUT file. Then, by defining two detection areas of equal size directly in front of and behind the absorber material, fluxes of primary and secondary gamma-rays were obtained. This operation is done with the F4 definition of the MCNPX code. Finally, a source with point isotropic behaviors was located at a point before the absorber glass material and the first detection area (see Figure 2). For each simulation cycle, the source description and hence the characteristic gamma-ray energy were determined. Meanwhile, 10⁸ particle counts were generated for each simulation cycle through random event generator of MCNPX. At the end of the simulation processes, the relative error rates from the OUTPUT file were observed to be less than 1%. MCNPX simulations were performed using the D00205ALLCP03 MCNPXDATA package is included of DLC-200/MCNPDATA cross-section libs (Figure 1).

Table 2

Radioisotopes and gamma-ray energies used for gamma-ray TF calculations

Radioisotope	Gamma-ray energy (MeV)
67Ga	0.0086, 0.0093, 0.1840
57Co	0.0144, 0.1221, 0.1365
111In	0.0230, 0.1710, 0.2450
133Ba	0.0532, 0.0796, 0.0810, 0.2764, 0.3029, 0.3560, 0.3838
201Tl	0.0710, 0.1350, 0.1670
99mTc	0.1405
51Cr	0.3201
131I	0.2843, 0.3645, 0.6370, 0.7229
58Co	0.5110, 0.8108
137Cs	0.6617
60Co	1.1732, 1.3325

Figure 2

(a) 2-D view of designed MCNPX simulation setup. (b) 3-D illustration of designed MCNPX setup (2-D and 3-D views are obtained from MCNPX Visual Editor VisedX22S).

3 Results and discussions

The TF values of the fabricated glass materials were calculated for 0.5, 1.5, 2.5, and 3 cm thicknesses as a function of the gamma-ray values emitted from different radioisotopes. Figure 3 depicts the TFs of investigated glasses and steel-magnetite concrete as a function of used radioisotope energy (MeV) at different glass thicknesses. As can be seen from the figure, the amount of the transmitted gamma-rays that all for glass samples and the conventional shielding materials (i.e., reinforced concrete) increased depending on the increasing radioisotope energy at all material thicknesses. Behaviorally, the common attitudes of all materials is the low transmission rate observed at low gamma-ray energies. This is due to the limited penetrating capability of low energy gamma-rays. On the other hand, for the same energy value, a certain TF value difference was observed proportionately between the material with the lowest thickness and the material with the highest thickness. This may be explained by the fact that the primary gamma-ray that penetrates the thin material does not undergo complete energy absorption. On the other hand, obtaining the lowest TF value at 3 cm, which is the thickest value for the same energy value, can be explained by the primary gamma-ray’s first interaction and subsequent interactions in the material occurring at a sufficiently large thickness and simply left the material with minimal quantitative values. As a consequence, the TF values for the radioisotope energies utilized in the glass samples and the concrete material used as the reference material were lowest at low energies and high thicknesses. In the largest context, this situation is linked to the penetrating qualities of gamma-rays [19,20,21,22], and no abnormal behavior tendencies have been identified. As the second step of TF value assessment, the TF values of glass and concrete samples were compared for identical thickness values, and the material with the lowest TF value and hence the best absorption characteristics was determined. Figure 4 depicts the comparison of the TFs as a function of used radioisotope energy (MeV) for different glass thicknesses and concrete. Figure 4 also quantitatively depicts the particular gamma-ray TF behavior of all materials investigated and compared at the same thickness. The TF value of the C series glasses fabricated according to the data obtained in this phase of the investigation is proportional to the ratio of CdO added to the glass at a certain thickness. Among the glass samples developed, the C40 sample with the greatest CdO doping ratio and densest structure exhibited the lowest TF values throughout the thickness range. This may be attributed to the positive contribution of the maximum CdO solid’s density rise to the gamma-ray absorption qualities. Meanwhile, the C40 sample’s TF values are lower than those of the concrete with additives. This demonstrates that these next-generation materials, which outperform the absorption capabilities of currently used heavy-mixed concrete, may be employed for identical reasons. In the last comparison stage of the study, the half value layer (HVL) values of all C series glasses were compared with the reinforced concrete material. Figure 5 depicts the variations of HVL (cm) with photon energy (MeV) for all C0–C8 glasses and concrete. A material’s HVL value may be defined for a given photon energy [23]. The HVL value is the material thickness necessary to quantitatively lower the intensity of photons at the relevant energy using that material [24,25,26,27,28,29,30,31,32,33,34,35]. Thus, the fact that the HVL value is low for a given photon energy value is a significant pattern of the material’s exceptional absorption capabilities. As seen in the figure, the C40 sample had the lowest possible HVL values at all energy levels. As a result, employing C40 samples instead of traditional concrete shields may be advantageous in terms of physical space requirements and space costs.

Figure 3

TFs of investigated glasses and steel-magnetite concrete as a function of used radioisotope energy (MeV) at different glass thicknesses.

Figure 4

Comparison of the TFs as a function of used radioisotope energy (MeV) for different glass thicknesses and concrete.

Figure 5

Variations of half value layer (cm) with photon energy (MeV) for all C0–C8 glasses and concrete.

4 Conclusion

One of the challenges that researchers have focused on in recent years is developing alternative, ecologically acceptable, nontoxic, and low-cost materials to substitute Pb and concrete in equipment and structural designs used in medical and industrial radiation applications. Currently, some specialized glasses are employed in a variety of configurations for a variety of purposes in nuclear and diagnostic radiology facilities. Our primary goal in this research was to observe the absorption differences across a broad radioisotope energy range caused by structural changes in the glass composition accompanied by certain chemical modifications and to investigate the relationship between these differences and the chemical modifications. The TF values of glasses synthesized utilizing the distinctive gamma-ray energies of a variety of therapeutic, diagnostic, and research radioisotopes, such as Barium, were determined for this purpose. The findings indicate that the linearly increasing CdO ratio in such a glass composition contributes positively to all gamma-ray energies. As a consequence, the C40 sample containing 40% CdO was shown to have the best absorption characteristics among the heavy and thick glasses prepared. More importantly, the gamma-ray absorption properties of this high-density glass manufactured were shown to be superior to those of conventional and heavy concrete sample. The critical point is that the transparency of the C40 sample, which exhibits superior qualities to conventional materials, allows for viewing of the radioactive source and prompt identification of any contamination that may occur, allowing required measures to be taken. Nevertheless, implementing the outcomes of this research in practice and expanding the allowable CdO additive quantity to greater levels may be a few critical future studies that may be recommended to scientific community.