Radiation Protection Research Articles

Construction of new polygon mesh-type phantoms based on adult Japanese voxel phantoms.

To construct adult Japanese phantoms applicable to individual exposure dose assessments, we created adult Japanese polygon mesh-type male (JPM) and female (JPF) phantoms through modification of the adult Japanese voxel phantoms, JM-103 (male) and JF-103 (female). The body sizes and masses of organs, tissues, and organ contents in JPM and JPF were adjusted to the Japanese averages, except for those unimportant for radiation protection or risks. The JPM and JPF data were converted to tetrahedral mesh-type data and incorporated into the Particle and Heavy Ion Transport code System (PHITS) for dose calculations. The dosimetric characteristics of the JPM and JPF phantoms were validated by calculating their effective doses in the anterior-posterior geometry for the external irradiation of photons with energies of 0.01-20 MeV and compared with those of JM-103 and JF-103 or the reference values given in ICRP Publication 116. The results confirmed no problems applying JPM and JPF to dose assessments in adult Japanese subjects. Furthermore, it was found that JPM and JPF can also accurately calculate the absorbed doses for entire organs and high radiosensitive cell regions with thin, small, and complicated structures.

Radiation safety for the RAON accelerator facility

Radiation safety for the RAON accelerator facility

Immunomodulatory hydrogel patches loaded with curcumin and tannic acid assembled nanoparticles for radiation dermatitis repair and radioprotection

Radiation dermatitis, as a major side effect of radiotherapy, is a form of skin injury linked to the production of reactive oxygen species (ROS). Development of functional patches with excellent ROS scavenging ability and biosafety is crucial for radioactive dermatitis repair. Herein, we present a nanohybrid hydrogel patch composed with natural-derived substances. The patch possesses dual functionalities of chemically scavenging ROS and physically absorbing radiation, thereby alleviating radiation damage. Curcumin@tannic acid (Cur@TA) nanoparticles are synthesized through spontaneous coordination of TA and iron ions (Fe3+), enabling the encapsulation of Cur. The nanoparticles are then loaded into gelatin methacryloyl (GelMA) hydrogel patches, which endow the hydrogel with tissue adhesion as well as photothermal properties. Simultaneously, the patch exhibits powerful ROS scavenging capability and promotes M2 polarization of macrophages. Additionally, the high-water content of the GelMA hydrogel enables physical absorption of low-energy X-rays, providing radiation shielding. Based on these characteristics, we have demonstrated the augmented therapeutic efficacy of the hydrogel patch loaded with Cur@TA nanoparticles in the mouse model of radiation dermatitis. Our findings introduce a novel therapeutic approach for radiation protection and treatment, with potential clinical applications.

German radiation oncology's next generation: a web-based survey of young biologists, medical physicists, and physicians-from problems to solutions.

Radiation science is of utmost significance not only due to its growing importance for clinical use, but also in everyday life such as in radiation protection questions. The expected increase in cancer incidence due to an aging population combined with technical advancements further implicates this importance and results in ahigher need for sufficient highly educated and motivated personnel. Thus, factors preventing young scientists and medical personnel from entering or remaining in the field need to be identified. Aweb-based questionnaire with one general and three occupation-specific questionnaires for physicians, biologists, and medical physicists working in radiation oncology and research was developed and circulated for 6weeks. While the overall satisfaction of the 218 participants was quite high, there are some points that still need to be addressed in order to ensure acontinuing supply of qualified personnel. Among these were economic pressure, work-life balance, work contracts, protected research time, and ademand for an improved curriculum. Mentoring programs, improved education, and strengthening the value of societies in radiation sciences as well as translational approaches and more flexible working arrangements might ensure ahigh-quality workforce and thus patient care in the future.

Research Progress on the Biological Activity of Ganoderic Acids in Ganoderma lucidum over the Last Five Years.

Ganoderma lucidum (G. lucidum) is a traditional edible and medicinal mushroom in China. The main bioactive components in G. lucidum include triterpenoids, polysaccharides, steroids, and sterols. Ganoderic acids (GAs) are one of the most abundant triterpenoids found in G. lucidum, garnering significant attention from researchers in the fields of medicine and health care. We summarize the extensive studies on the physiological function of GAs in anti-cancer, anti-inflammatory, radiation protection, anti-aging, liver protection, anti-microbial, and neuroprotection areas, among others. This review provides a comprehensive overview of the recent advances in the bioactivities and pharmacological mechanisms of GAs, aiming to delineate the current research directions and the state of the art in this field. This analysis helps to rapidly identify new bioactivities of GAs and understand their mechanisms, leading to more effective treatments for various diseases.

Unveiling the shielding potential: Exploring photon and neutron attenuation in a novel lead-free XCr2Se4 chalcogenide Spinals alloys with MCNP 4C code

The study covers shielding properties of three promising lead-free XCr2Se4 alloys, where X represents Mn, Cu, or Ag, across a spectrum of energies between 0.02 MeV and 15 MeV. MCNP4C was employed to simulate the mass attenuation coefficient (MAC) for the three selected alloys. After being prepared using the conventional solid-state reaction method, the samples displayed densities ranging from 5.45 g cm−3to 6.22 g cm−3. The simulated results obtained from the MCNP4C are then benchmarked with data obtained using different software such as Phy-X, Py-MLBUF, EPIXS, and GRASP. Upon comparing the acquired results with the Phy-X data, a notable correlation was observed with a relative difference ranging from 0.11% to 8.57%. The accuracy of the simulated data was additionally validated through the Kolmogorov-Smirnov test. From the simulated MAC, different ionizing radiation shielding properties including linear attenuation coefficient (LAC), half-value layer (HVL), mean-free path (MFP), transmission factor (TF), radiation protection efficiency (RPE), and fast neutron removal cross section (FNRCS) were calculated to provide a comprehensive analysis of the samples' efficacy in dealing with different types of radiation. The LAC for neutrons at different energies is also examined. HVL analysis indicates the radiation-blocking capacity of the samples. The TF shows the alloys’ ability to either permit or restrain the transfer of radiation. Additionally, RPE and FNRCS give indication of the shielding potential of the studied samples. This study is important for nuclear physicians and researchers and serves as useful data for the development of superior shielding materials.

Thyroid scintigraphy of healthy cats using small-field-of-view gamma cameras.

Small-field-of-view (SFOV) gamma cameras can offer higher sensitivities than conventional gamma cameras. However, there are currently no reports on the efficacy and safety of thyroid scintigraphy using SFOV gamma cameras in veterinary medicine. Therefore, we aimed to evaluate the efficacy and radiation safety of an SFOV gamma camera for feline thyroid scintigraphy. Three veterinary staff members (operator, staff 1, and staff 2) performed thyroid scintigraphy on 10 healthy cats in this study. The operator administered either 2 or 4 mCi of technetium-99m pertechnetate (99mTcO- 4) through the cephalic vein. At 20, 40, and 60 min after injection, thyroid images were obtained using a SFOV gamma camera under various acquisition conditions (100,000, 150,000, and 200,000 counts and 30 and 60 s). Thyroid scintigraphy images were analyzed by calculating the thyroid-to-salivary ratios (TSR) and thyroid-to-background ratios (TBR). Surface and ambient radiation were measured hourly from immediately after injection to 6 h. The cumulative occupational radiation doses were measured during the procedure. The TSR and TBR median values aligned with the previously reported normal range obtained using a large-field-of-view gamma camera. There were no notable differences in TSR and TBR between the two doses of 99mTcO- 4, nor across acquisition conditions and timelines. The 4-mCi group consistently emitted more ambient (p < 0.05) and surface (p < 0.05) radiation than did the 2-mCi group. Staff 1 consistently received higher cumulative radiation doses than did staff 2 and the operator (p < 0.05). The SFOV gamma camera demonstrated adequate image quality for thyroid scintigraphy in healthy cats even with relatively low doses and short acquisition conditions. Radiation exposure during the procedure posed minimal safety concerns. Therefore, the SFOV gamma camera could be a valuable tool for evaluating thyroid glands in cats.

2. Measurement Division, Diagnostic Imaging Division, Medical Informatics Division, Radiation Protection Division

2. Measurement Division, Diagnostic Imaging Division, Medical Informatics Division, Radiation Protection Division

Analysis of personal dose monitoring results of occupational external exposure for radiation workers in a hospital

Objective: To analyze and understand the level of occupational external radiation personal dose of radiation workers in a hospital, and provide a reference for radiation protection. Methods: In May 2023, the personal dose monitoring data of all radiation workers in the hospital from 2018 to 2022 were collected, and their occupational external radiation exposure levels were analyzed and evaluated. All radiation workers were divided into radiation diagnosis, radiotherapy, interventional radiology, and nuclear medicine according to their occupational categories. Results: A total of 1626 person-times were monitored from 2018 to 2022, and the average annual effective dose per person (0.13-0.36 mSV) was lower than the national limit of 20 mSv. The average annual dose of nuclear medicine was the highest (0.36 mSv), and interventional radiology was lower (0.13 mSv). The average annual dose of individuals was the highest in 2020 (0.24 mSv) . Conclusion: The average annual effective dose per person of radiation workers in this hospital is lower than the national standard, and it is still necessary to strengthen the inspection of the wearing of personal dosimeters and the using of protective equipment for radiation workers.

Research on the radioactive safety of industrial steam supplied by PWR nuclear power plant

The heating source of the nuclear power steam supply installation is PWR secondary steam, but the secondary steam is potentially radioactive, and the radioactive material may migrate to the industrial steam, which goes through the leakage of the heat exchanger. Compared with the conventional industrial steam, the steam supplied by the nuclear power has potential radioactivity. Nuclear power is first introduced to produce the industrial steam, in order to analyze the specific activity and radiation effects of the industrial steam on the staffs at the user side, an analytical model of industrial steam radioactivity is developed based on the technological process of industrial steam under operating conditions, and then the exposure pathways of the staff at the user side from the industrial steam is analyzed. The specific activity of the industrial steam and the exposure dose of the staff is evaluated. According to the evaluation results, it can be seen that the dose to staff caused by the industrial steam is far smaller than the radiation impact on the public caused by the nuclear power plant, and can meet the requirements of relevant regulations and standards. Therefore from the radiation safety point of view, the industrial steam supplied by the nuclear power steam supply installation is safe and acceptable for industrial use.

Enhanced optical and structural traits of irradiated lead borate glasses via Ce3+ and Dy3+ ions with studying Radiation shielding performance

Lead borate glass is the best radiation shielding glass when lead is in high concentration. However, it has low transparency after radiation exposure. Radiation decreases transparency due to chemical and physical changes in the glass matrix, such as creating or healing defects in the glass network. The addition of rare earth elements like cerium and dysprosium oxides to lead borate glasses can improve their transparency and durability as radiation shielding barriers. The newly manufactured glasses’ optical absorption, structural, and radiation shielding properties were measured. The optical characteristics of the generated samples were examined to determine the effect of the cerium/dysprosium ratio on the structural alterations, specifically in the presence of bridging oxygen (BO) and non-bridging oxygen (NBO). Incorporating Ce3+ results in peaks at 195 nm for borate units, 225 nm for Ce3+, and a broadened peak at 393 nm due to overlapping peaks for Ce3+ and Ce4+ in the UV region. By adding Dy, multiple peaks are observed at 825, 902, 1095, 1275, and 1684 nm, corresponding to the transition from 6H15/2 ground state to 6F5/2, 6F7/2, 6F9/2, 6F11/2, and 6H11/2. The samples were also tested before and after exposure to gamma irradiation from a 60Co source at a dose of 75 kGy to assess their stability against radiation. The energy gap value during irradiation shows decreased non-bridging oxygen. The energy gap difference before and after irradiation for the M4 sample shows higher NBO to BO conversion, reducing radiation damage and improving structural stability. Furthermore, X-ray photoelectron spectroscopy was utilized to get insight into the coordination chemistry of the created glass samples. The half-value layer (HVL), radiation protection efficiency (RPE), neutron removal cross-section (FRNCS), mean free path (MFP), mass attenuation coefficients (MAC), and effective atomic numbers (Zef) of the glassy structure were calculated theoretically to assess its radiation shielding qualities. The linear attenuation coefficient order for the prepared samples was M1 > M2 > M3 > M4. The FRNCS values were 0.090, 0.083, 0.081, and 0.079 cm−1 for samples M1, M2, M3, and M4, respectively.

Evaluation of radiation protection knowledge and practices among Moroccan operating room nurses

Evaluation of radiation protection knowledge and practices among Moroccan operating room nurses

Low doses of ionising radiation: definitions and contexts

The term 'low dose' is applied to different levels of dose depending on the circumstances of exposure, with the potential for confusion unless the reasoning is clear. The United Nations Scientific Committee on the Effects of Ionising Radiation has defined low absorbed doses of ionising radiation as below about 100 mGy, and low dose rates as below 0.1 mGy min-1 (6 mGy h-1). These values relate to the interpretation of scientific evidence from epidemiological and biological studies. The International Commission on Radiological Protection has used similar values of 100 mSv and 5 mSv h-1 and applied this categorisation directly to the specific situation of patients undergoing diagnostic procedures: doses below 100 mSv were referred to as 'low' and doses below 10 mSv as 'very low'. Consideration of other exposure situations suggest that the same terms can be used for exposures received by emergency workers. However, for workers and members of the public in planned exposure situations, it is suggested that the term 'low dose' applies to doses below 10 mSv and 1 mSv, respectively - that is, below the dose limits. In each case, dose is being used as a surrogate for risk - risks at low doses are uncertain and estimates may change, but order of magnitude considerations are sufficient in most cases. Doses of < 100 mSv, < 10 mSv and < 1 mSv correspond to life-time cancer risk estimates of the order of < 10-2, < 10-3 and < 10-4, respectively.&#xD.

Thermal cross section of poly(2-hydroxyethyl methacrylate) hydrogels for neutron dose calculation

Phantom materials are used to design radiation safety and protection strategies by means of numerical dose calculation. In the case of thermal neutrons, the radiation transport in the phantom relies on mass attenuation coefficients and total cross sections which are dependent on the physical and chemical properties of the material. Specifically for medical applications, such as neutron capture therapy, the neutron-induced dose is mainly related to the neutron absorption by hydrogen and nitrogen in the human tissue and body. Here, we investigate the use of poly(2-hydroxyethyl methacrylate) as a phantom material by experimental measurement and modeling of its total neutron scattering cross section and mass attenuation coefficient. We show that by varying the hydration level from 10 to 40 w%, one can obtain a neutron attenuation coefficient similar to polymethyl methacrylate or more representative of the human body, respectively. By benchmarking our phenomenological model on the experimental data, we provide a new example of how to use the average functional group approximation to accurately model total neutron cross sections in the framework of personalized medicine approaches.

Machine learning-enhanced stochastic uncertainty and sensitivity analysis of the ICRP human respiratory tract model for an inhaled radionuclide

The International Commission on Radiological Protection (ICRP) has developed the reference Human Respiratory Tract Model (HRTM), detailed in ICRP Publications 66 and 130, to estimate the deposition and clearance of inhaled radionuclides. These models utilize reference anatomical and physiological parameters for particle deposition (PD). Biokinetic models further estimate retention and excretion of internalized particulates, aiding the derivation of inhalation dose coefficients (DC). This study aimed to assess variability in deterministic131I biokinetic and dosimetry models through stochastic analysis using the updated HRTM from ICRP Publication 130. The complexities of the ICRP PD model were reconstructed into a new, independent computational model. Comparison with reference data for total PD fractions for reference worker, solely a nose breather, covering activity median aerodynamic diameters from 0.3μm to 20μm, showed a 1.04% relative and 0.7% absolute difference, demonstrating good agreement with ICRP deposition fractions. The deterministic DC module was reconstructed in Python and expanded for stochastic analysis, systematically expanding deposition components from HRTM and assigning probability distribution functions to uncertain parameters. These were integrated into an in-house stochastic radiological exposure dose calculator, utilizing latin hypercube sampling. A case of an occupational radionuclide intake was explored, in which biodistribution and committed effective DC (CEDC) were computed for131I type F, considering a lognormal particle size distribution with a median of 5μm. Results showed the published ICRP reference CEDC marginally exceeds the 75th percentile of observed samples, with log-gamma distribution as the best-fit probability distribution. A Random Forest regression model with SHapley Additive exPlanations was employed for sensitivity analysis to predict feature importance. The analysis identified the HRTM particle transport rates scaling factor, followed by the aerodynamic deposition efficiency in the alveolar interstitial region as the most impactful parameters. This study offers a unique stochastic approach on inhaled particulate metabolism, enhancing radiation consequence management, medical countermeasures, and dose reconstruction for epidemiological studies.

Neutron spectrometry of a 241Americium-Boron neutron source using the NCT-WES single-moderator neutron spectrometer

241Americium-boron (α,n) neutron sources have been produced for various application from nuclear industry to well logging or radiation protection. Compared to 241Americium–beryllium sources their specific emission rate is lower, but their spectrum is narrower, and their production cycle uses boron, which is less toxic than beryllium. Very few data are available in literature about the energy distribution of this neutron source: the 2001 version of Standard ISO 8529-1 reported a reference spectrum derived from 1970s data, exhibiting a single peak from about 1 to 6 MeV. Other spectra are available in recent works from PTB and NPL, based on high-resolution spectrometers and Bonner spheres. ENEA Frascati owns a 241Am-B neutron source with nominal emission rate 3.5 × 106 s−1. Knowing its spectrum is important, as this source is used to feed the HOTNES (Homogeneous Thermal Neutron Source) facility. A spectrometry experiment was organized relying on the recently developed NCT-WES neutron spectrometer. Belonging to the family of the Single Moderator Neutron Spectrometers, NCT-WES is a convenient alternative to Bonner spheres as it derives the whole spectrum from a single exposure. The experimental data were elaborated in comparison with the existing literature spectra. As a main results of the study, the spectrum of the ENEA 241Am-B neutron source nearly perfectly agrees with that derived at NPL.

Production and evaluation of the polyester composites containing pyrite and niobium diboride for radiation protection

The presented study focused on investigating the radiation shielding properties of polyester/pyrite/niobium diboride (FeSNbB) composites. Various gamma-ray shielding parameters were evaluated with help of a HPGe detector system. Theoretical and simulation results are further evaluated by WinXCOM, GEANT4 and FLUKA codes, and the obtained results are compared with experimental values in the energy region 59.5 keV–1332.5 keV. The prepared samples are additionally evaluated using kerma relative to air values and Exposure build-up factors. Then, for the evaluation of neutron radiation shielding, effective removal cross sections for fast neutrons were determined for FeSNbB composites. Electron attenuation efficiency (AE%) for the produced composites was assessed across the energy spectrum of 4 MeV–18 MeV, utilizing experimental data from a Linac machine and theoretical predictions from Eclipse TPS. The results were found to be compatible with each other. FeSNbB50 has the best shielding properties among the studied composites.

Exploring the diverse modifiers influence on the structural, physical, and mechanical properties of a zinc oxide-based boro-tellurite glass system for enhancing radiation shielding performance

This work is meant to fabricate zinc boro-tellurite (ZnO–TeO2–B2O3) glass systems with various modifiers, including MoO3, Bi2O3, and PbO, using the melt quench technique. Four glass samples were produced, each comprising 20 % of a specific modifier from the total weight percent of the neat glass system. Each glass sample is given a code: TBZn, TBZMo, TBZBi, and TBZPb. The XRD results showed an amorphous nature for all glass samples. The FTIR results showed the functional groups for borate and tellurite oxide. The mechanical properties were reduced when different oxides were added instead of ZnO. For example, Young's modulus decreased for glass samples; their recorded values for Young's modulus are 74.516, 72.569, 60.526, and 59.272 GPa for TBZn, TBZMo, TBZBi, and TBZPb, respectively. On the other hand, adding Bi2O3 and PbO led to enhanced shielding properties. For instance, the linear attenuation coefficient values for TBZn, TBZMo, TBZBi, and TBZPb at 0.015 MeV are 168.739, 115.656, 380.711, and 279.162, respectively. Also, the radiation protection efficiency (RPE) at 5 cm and 600 keV is 83.220 %, 79.299 %, 93.919 %, and 90.222 % for TBZn, TBZMo, TBZBi, and TBZPb, respectively. While adding different oxides decreased stability in the current glasses, it also enhanced their shielding properties, recommending their use in the radiation shielding field.

Assessing the Level of Radiation Protection and Safety Compliance of Radiographers Currently Employed in the Makkah Cluster Hospitals

Assessing the Level of Radiation Protection and Safety Compliance of Radiographers Currently Employed in the Makkah Cluster Hospitals

Research on B4C/PEEK Composite Material Radiation Shielding

There are various types of charged particles in the space environment, which can cause different types of radiation damage to materials and devices, leading to on-orbit failures and even accidents for spacecraft. Developing lightweight and efficient radiation-shielding materials is an effective approach to improving the inherent protection of spacecraft. The protective performance of different materials against proton and electron spectra in the Earth’s radiation belts is evaluated using a Geant4 simulation. Based on the simulation results, suitable hardening components were selected to design composite materials, and B4C/PEEK composites with different B4C contents were successfully prepared. The experimental results demonstrate that the simulated and experimental results for the electron, proton and neutron shielding performance of the B4C/PEEK composites are consistent. These composites exhibit excellent radiation shielding capabilities against electrons, protons and neutrons, and the radiation protection performance improves with increasing B4C content in the B4C/PEEK composite materials.