Optimization study on neutron/γ radiation–protective clothing materials with computational human phantoms

Abstract

The applications of nuclear science and technology in both production and daily life are becoming increasingly widespread. Radiation shielding, as a critical component, ensures environmental safety and protects human health. In this study, 20 shielding schemes were designed using ethylene–propylene diene monomer as the base material. These schemes incorporated various proportions of boron carbide and gadolinium oxide as neutron-absorbing components and tungsten as the gamma-shielding component. Based on the Chinese reference adult male (CRAM) voxel model and using an anterior-posterior (AP) irradiation setup, the Monte Carlo method was employed to calculate 28 organ/tissue doses and effective dose reductions from neutron and gamma radiation across the 20 material compositions. Each case was evaluated at three different thicknesses—1, 3, and 5 mm—with Monte Carlo calculation errors controlled within 1%. Results indicated that, for any composite shielding material, the 5-mm thickness provided optimal protection. When an unmoderated and unthermalized ²⁵²Cf neutron source was used, effective dose reductions ranged from 32.60% to 38.75% compared to the unshielded case. With a monoenergetic neutron source at 1 keV, the reduction range was between 57.62% and 69.42%. The trend in changes for different composite shielding materials under neutron sources at different energy levels is consistent. When ¹³⁷Cs served as the gamma source, effective dose reductions ranged from 7.96% to 20.97%, demonstrating that the composite materials offer substantial protection for both neutron and gamma radiation. Additionally, it was found that organs partially exposed outside the shielding material experienced a slight increase in dose due to neutron scattering. In practical applications, full-body protection should be implemented to mitigate this issue.

Introduction

The continuous increase in energy demand and the severe challenges posed by climate change are driving the unprecedented development and application of nuclear power, a mature technology that is both low-carbon and environmentally friendly. Nuclear safety assurance has been established as a critical strategic task for national security and as a key responsibility toward individuals and the environment. During reactor emergency repairs or routine equipment maintenance, effective personnel protection measures must be taken to avoid the impact of high neutron radiation level on the health. Additionally, radiation protection suits are required in various neutron application scenarios, including aerospace, radiology/nuclear medicine treatment, irradiation processing, neutron flaw detection, and scientific research. Personal protective equipment or devices specifically designed for aviation applications have yet to be developed [1]. Moreover, neutrons, compared to alpha and beta particles, have stronger penetrating power and can cause more damage to the human body [2]. Therefore, it is urgent to develop new n/γ composite radiation–protective materials that enhance shielding effectiveness while improving wearer comfort.

Some studies have been conducted to develop neutron shielding materials. Xu et al. [3] designed a double-layer shielding material consisting of tungsten–nickel alloy and rare earth polymer materials, but it is a hard material designed for ships. There are also shielding materials designed for radiation therapy rooms [4] and spacecraft [5]. And explorations into new shielding materials such as aluminum base materials [6] and ceramic materials [7].

He et al. [8] designed a three-layer shielding material containing tungsten rubber and gadolinium rubber, investigating the shielding rate for neutron and gamma-ray particle intensity. The material exhibited a shielding rate of 21.52% for medium-energy gamma rays and 78.7% for slow neutrons. Wang et al. [9] developed a protective material using ethylene–propylene diene monomer (EPDM) as a base material mixed with B₄C, and the shielding efficiency of 1.5-cm material was 42.3%. Özdemir et al. [10] calculated the shielding rate by incorporating boron trioxide into EPDM, and the particle flux of 2-mm material was reduced by 20%. Abdel-Aziz et al. [11] added 47% B₄C to EPDM and calculated the reduction rate of slow neutron flux. Güngör et al. [12] introduced hexagonal boron nitride into EPDM, achieving a thermal neutron attenuation rate of 61.5%. Huo et al. [13] prepared polyethylene materials doped with Gd₂O₃ and B₄C, enhancing the shielding rates for neutrons and gamma rays. Kim et al. [14] designed a nanofiber fabric for flight attendants by incorporating Gd₂O₃ and W into polyurethane; the 0.3-mm thick material had shielding rates of 17.5% and 15.2% for neutrons and photons, respectively. However, current research on integrated flexible composite shielding materials for medium-energy neutrons and gamma rays is scarce and there are no studies calculating the dose of human organs after shielding with new materials, with most evaluations being based on particle flux calculations to assess shielding rates.

Based on the above background and current research status, the shielding effectiveness of radiation-protective materials with varying thicknesses (1, 3, and 5 mm) and doping compositions was investigated using the Monte Carlo calculation method. These materials use EPDM as the base material and are mixed with Gd₂O₃, B₄C, and W in 20 different ratios. By calculating the organ doses and effective doses of the phantom after shielding with these materials, the study evaluates their protective effectiveness in radiation protection. Material selection and data support are thus provided for radiation-protective clothing research. Additionally, the integrated materials ensure uniform protection performance and the flexible materials maximize the comfort of the personnel.

Materials and methods

Description of composite shielding materials

Neutrons, as uncharged neutral particles, possess strong penetrating abilities. Neutron shielding requires moderating and absorption. The moderating of neutrons relies mainly on light nuclei (such as hydrogen, carbon, and oxygen), while absorption mainly uses elements with large absorption cross sections.

EPDM has a high hydrogen content and can effectively slow down neutrons [15]. Moreover, it is rich in softness and elasticity, so it is a suitable substrate for shielding. The main neutron-absorbing elements with large absorption cross sections [16] include ¹⁰B, ¹¹³Cd, ¹⁴⁹Sm, and ¹⁵⁷Gd. Cd [17] has high biological toxicity, while Sm [18] has inherent radioactivity. Therefore, boron and gadolinium were selected as neutron-absorbing components in this study. The absorption cross sections and abundances of the two are shown in Table 1. As indicated in the table, the thermal neutron absorption cross sections of ¹⁵⁷Gd and ¹⁵⁵Gd are 66 times and 16 times of ¹⁰B and the abundances of the two stable isotopes are as high as 30%. At the same time, due to the high frequency number of gadolinium, it also has a certain absorption effect on gamma rays.

Tungsten has better shielding performance than lead and does not contain biological toxicity; it is an ideal γ-absorbing component and can also slow down the neutron. At the same time, the absorption edge of the K layer of gadolinium can compensate for the weak absorption area of tungsten [8] and the combination of the two can bring better shielding effectiveness.

In summary, this study selected EPDM as the base material, incorporating Gd₂O₃ in mass fractions up to 15 wt%, B₄C between 10 wt% and 15 wt% mass fraction and tungsten powder up to 55 wt%, with increments of 5 wt%. This process yielded 20 variations of composite shielding materials with different functional filler ratios. The mass fraction of EPDM was progressively reduced in each case, and the material composition for each scenario is listed in Table 2.

The effect of Gd₂O₃, B₄C, and W on neutron and γ shielding can be evaluated by comparing each pair of cases.

The thickness of the material used in the manufacture of personal protective equipment is between 1 and 5 mm [8, 9, 14, 19, 20]. To account for wearer mobility and to comprehensively assess the effect of thickness on shielding performance, this study calculates the dose reduction achieved with materials of three different thicknesses—1, 3, and 5 mm—at intervals of 2 mm.

Voxel phantom

The voxel phantom used is Chinese reference adult male (CRAM) voxel model [21]. The voxel phantom is a high-resolution whole body element model, containing almost all the organs recommended in International Commission on Radiological Protection (ICRP) Publication 103 [22], and the organ quality meets the Chinese reference data. The spatial distribution of the details of bone composition is also considered. The voxel size of the body model is 1.741 × 1.741 × 1 mm, and the matrix size is 276 × 156 × 1700. The use of the voxel body model can ensure that the calculated organ dose is more in line with the Chinese constitution.

Monte Carlo simulation method

The Monte Carlo particle transport code used in this study is THUDose_PD [23], a 3D radiation dose calculation and protection design simulation program developed by our research team. This program can support the transmission of electrons, photons, neutrons, protons, and heavy ions and realize the multi-threaded parallel computation of central processing unit (CPU) and graphics processing unit (GPU). The ENDF/B-VI database is used internally. The program has been widely verified [24].

A simplified protective suit was modeled as an elliptical cylinder covering the trunk portion of the CRAM from the thyroid to the testicles. The elliptical cross-section of the internal cylinder has a major axis of 33.1 cm and a minor axis of 14.3 cm, with a height of 76.4 cm. The parameters of the elliptic cylinder are calculated by MATLAB R2022B after reading the voxel phantom to ensure that the chest and abdomen can be fully enclosed. Considering that neutrons have a significantly higher radiation weighting factor—typically 2.5–20 times greater than that of gamma rays [22]—their biological impact on the human body is notably more pronounced. The shielding effectiveness of the composite material against neutron sources was first calculated, followed by separate calculations for photon sources. The source terms for neutrons and gamma are ²⁵²Cf fission neutron sources [25] and ¹³⁷Cs, respectively. Since the energy of neutrons from a ²⁵²Cf fission source is ~2 MeV, moderation materials are commonly added near the neutron source in studies to thermalize the neutrons. Given the unique geometry of the AP irradiation and the phantom model, this study further evaluates the shielding effectiveness of composite materials for a monoenergetic neutron source at 1 keV. The irradiation orientation is AP irradiation. To ensure the statistical error is <1%, the number of particles is 1 × 10⁸. The diagram is shown in Fig. 1.

Evaluation of shielding effectiveness

THUDose_PD was used to calculate the organ dose (28 organs/tissues including the heart, liver, kidney, etc.) and the effective dose under the 20 different materials shielding. The units are normalized to the dose per particle (pSv/ptc). The calculated results are compared with the organ dose and effective dose without any shielding state, which can be used as the evaluation standard to evaluate the protective effectiveness of shielding material against radiation. The relative difference is calculated as follows:

where D represents the organ dose/effective dose after the shielding and D₀ represents the organ dose/effective dose of the organ without shielding.

Results and discussion

The influence of material thickness on shielding effectiveness

The organ dose of the same composite shielding material under three different thickness shielding was compared to evaluate the effect of thickness. The lungs, kidney, heart, and red bone marrow (RBM) were treated as the organs at risk. As shown in Fig. 2, the presence of any composite material reduces the organ dose from radiation compared to the unshielded case. For the same composite shielding material, the organ dose decreased significantly with the increase of material thickness. Therefore, the 5-mm thickness provides the most effective shielding.

All organ doses in Fig. 2 strictly follow the rule of ‘the greater the thickness of the shielding material, the smaller the organ dose’. Although the trend of some cases of RBM is not very significant. This is because the organ dose of some RBM not located in the chest and abdomen is not affected by the thickness of the shielding material.

In addition, the effective dose reduction rates η corresponding to the three thicknesses were compared and the results were shown in Fig. 3. Each of the thickness distributions is the result of 20 different cases.

Figure 3 illustrates that with increasing thickness of the material, η increases, which means the effective dose decreases more. Consequently, the shielding effectiveness of the material with a thickness of 5 mm is the most effective.

Influence of composition of composite shielding material on shielding effectiveness

According to the conclusion of the previous section, 5 mm is the best choice for composite shielding materials. Therefore, the normalized organ dose (pSv/ptc) of 8 organs/tissues, including the brain, liver, and lung, etc. under 5 mm composite shielding were plotted in Fig. 4.

It can be seen that except for the brain, the dose of all organs after the composite shielding was significantly lower than that without the shielding. Using the thyroid gland as a case study, the organ dose without shielding measures at 1.01E-3 pSv/ptc. Following the application of shielding, this value is reduced to ~6E-4 pSv/ptc. The effect of several composites on the brain is similar. Since the head was not obscured by the composite shielding material, the presence or absence of the composite shielding material had little effect on it.

Further, the attenuation rate of the effective dose under radiation from two neutron sources of different energies are illustrated in Fig. 5.

In the absence of any shielding composite materials, the effective dose registers at 6.69E-4 pSv/ptc. As shown in Fig. 5, subsequent to shielding, η ranges from 32.60% to 38.75%. For an unmoderated and unthermalized ²⁵²Cf neutron source, the effective dose reduction reaches up to 38.75%. In the case of a monoenergetic 1-keV neutron source, the effective dose reduction rate exceeds 60% in most cases, with only one exception (57.62%), reaching a maximum reduction of 69.42%. According to the shielding performance criteria proposed by He et al. [8]—specifically a 20% shielding rate for medium-energy gamma rays and a 60% shielding rate for neutrons (based on the flux attenuation rate of a ²⁵²Cf neutron source moderated with 5 cm of polyethylene)—the 5-mm composite shielding material for a 1-keV neutron source achieves the 60% shielding standard in all cases except for Case 3. A cross-comparison of different composite shielding materials with equivalent mass fractions of EPDM and B₄C, such as in Cases 1/2/3 or 4/5/6, indicates that the addition of small amounts of gadolinium oxide enhances the shielding performance, although excessive additions diminish the effect. Therefore, the inclusion of functional materials must be optimized rather than maximized; detailed simulations based on specific conditions are necessary to achieve optimal protection. Among all composite shielding materials, Cases 4 and 9 demonstrate the best shielding performance overall. For the ²⁵²Cf neutron source, Cases 4, 7, 9, and 10 show the most effective shielding performance.

To more comprehensively demonstrate the shielding effectiveness of the four optimal configurations against the ²⁵²Cf neutron source, Fig. 6 illustrates the η values for both organ dose and effective dose.

As can be seen from the Fig. 6, organ dose decreased significantly after shielding. Only three organs— oral mucosa, extrathoracic region, and salivary glands—show a slight increase. In all four cases, an increase in dose to the salivary glands is observed, attributed to their proximity to the upper edge of the shielding material, where neutron scattering leads to this dose increase. In Case 9, there is a notable increase in the oral mucosa dose compared to the other three configurations in Fig. 6. This increase is attributed to the higher tungsten mass fraction in Case 9 (45 wt%), as tungsten, with its high atomic number, enhances neutron scattering, resulting in a dose increase in nearby organs.

Combining the observations from Figs 5 and 6, it can be seen that the compositions in Case 4 (EPDM 50 wt%, Gd₂O₃ 5 wt%, B₄C 10wt%, and W 35 wt%) and Case 9 (EPDM 40 wt%, Gd₂O₃ 5 wt%, B₄C 10wt%, and W 45 wt%) can achieve the largest reductions in organ dose and effective dose. However, for neutron sources with different energies, the optimal composition varies accordingly. This finding also indicates that the enhancement of the shielding effectiveness is not merely a matter of accumulating active components but requires comprehensive simulations to optimize the composition.

After the thickness of the composite shielding material was determined to be 5 mm, the relative difference of the organ dose after ¹³⁷Cs shielding was obtained, as shown in Table 3.

As shown in Table 3, similar to the neutron source scenario, dose reductions are observed in all organs after shielding, except for the brain, salivary glands, and extrathoracic region, which experience slight dose increases due to their location outside the shielding material. The reduction rates range from 0.39% to 32.35%. The η value for the lungs, as calculated, ranges from 8.64% to 22.89%. Among the 20 composite shielding materials, Case 15 meets the 20% γ-ray attenuation standard, primarily due to its high tungsten content of 55 wt%. In practical applications, this material could be utilized to create full-body personal protective equipment.

Organ doses due to photon sources are generally an order of magnitude lower than neutron sources. For example, for ²⁵²Cf, the effective dose of the composite shielding of Case 4 is 4.12E-4, while the effective dose of ¹³⁷Cs is 8.52E-5. Therefore, when the shielding effectiveness of each composite shielding material for photons is similar, priority should be given to their shielding effectiveness for neutrons. Therefore, this study suggests that the composition of 5-mm materials with the compositions in Case 4 (EPDM 50 wt%, Gd₂O₃ 5 wt%, B₄C 10wt%, and W 35 wt%) and Case 9 (EPDM 40 wt%, Gd₂O₃ 5 wt%, B₄C 10wt%, and W 45 wt%) are the optimal choices. In Case 4, the dose reduction for organs/tissues exposed to the ²⁵²Cf neutron source ranges from 0.61% to 59.05%, while for Case 9, the reduction spans from 3.72% to 54.10%. For a 1-keV neutron source, Case 4 achieves dose reductions between 25.64% and 98.09% and Case 9 achieves reductions between 25.88% and 83.61%.

Conclusion

A novel doped composite shielding material was designed, and the shielding effectiveness for neutrons and γ-rays of various compositions was simulated and calculated. This material has been shown to provide effective protection against both neutron and γ radiation. Furthermore, it is demonstrated that the inclusion of Gd within a certain proportion significantly enhances the shielding effectiveness, while excessive amounts can adversely impact protective performance. Therefore, the proportions of each component must be carefully calculated to optimize shielding efficiency; simply increasing the amount does not necessarily yield better results. Among the nine composites with different ratios calculated in this study, Case 4 (EPDM 50 wt%, Gd₂O₃ 5 wt%, B₄C 10wt%, and W 35 wt%) and Case 9 (EPDM 40 wt%, Gd₂O₃ 5 wt%, B₄C 10wt%, and W 45 wt%) are the best choices. For ²⁵²Cf, the effective dose was reduced by 38.75% compared with no shielding; for a 1-keV neutron source, this reduction reaches 69.42%. Some organs/tissues exposed outside the shielding material have an increased dose due to neutron scattering, but this can be avoided in practical applications by adding a head and neck mask. This study can provide data and theoretical support for the design of integrated neutron/γ radiation–protective clothing materials.

Conflict of interest

None declared.

Funding

This study was supported by the National Natural Science Foundation of China (Grant No. U2167209, Grant No. 12375312 and Grant NO. U23B2067).

References