Search

A retrospective analysis of occupational exposure to ionizing radiation from medical uses and industrial uses in the three provinces of Central China from 2000 to 2021 was conducted. The average annual effective dose in medical uses and industrial uses decreased from 2.042 mSv and 2.334 mSv in 2000–2002 to 0.476 mSv and 0.371 mSv in 2021 respectively; the fraction of monitored workers receiving annual dose not exceeding 1 mSv increased from 60.78% and 74.45% in 2000–2002 to 94.20% and 96.85% in 2021 respectively, while receiving annual doses exceeding 20 mSv declined from 1.35% and 1.91% in 2000–2002 to 0.18% and 0.03% in 2021 respectively. The average annual effective dose and NR 20 in the period 2000–2021 were relatively high in professional public health institutions (0.955 mSv and 0.004) and hospitals (0.815 mSv and 0.004). In 2021, the average annual effective dose to monitored workers in different occupational categories in medical uses in the three provinces of Central China were in the range of 0.199–0.692 mSv, with interventional radiology received the highest dose and NR 20 (0.692 mSv and 0.005); the average annual effective dose ranged from 0.161 to 0.493 mSv in industrial uses, with industrial radiography received the highest dose and NR 20 (0.493 mSv and 0.001). Occupational exposure in medical uses and industrial uses declined obviously in Central China, and the groups receiving higher doses are the radiation workers working in hospitals and professional public health institutions, or engaged in interventional radiology, nuclear medicine and industrial radiography, warranting more effective radiation protection measures. • The occupational radiation dose to radiation workers declined obviously in Central China in the period 2000–2021. • Interventional radiology and nuclear medicine have a higher occupational exposure risk than other medical uses. • Industrial radiography has a high occupational exposure risk than other industrial uses. • Radiation workers in hospital and public health professional institutions are more likely to receive high radiation doses. [ABSTRACT FROM AUTHOR]

In this study, Ni–Cu–Zn Fe 2 O 4 ferrite nanoparticles have successfully been synthesized utilizing the Co-precipitation technique. The primary objectives encompassed elucidating phase purity, discerning functional groups, scrutinizing surface morphology, and conducting structural analyses. To accomplish these objectives, a battery of advanced characterization techniques was employed, including power X-ray diffraction, Transmission infrared spectroscopy, UV–Visible spectrophotometer, and Scanning electron microscopy. Furthermore, the investigation was extended to the assessment of the gamma ray shielding properties exhibited by the synthesized Ni–Cu–Zn Fe 2 O 4 nanoparticles, spanning an energy range from 122 keV to 1330 keV. This evaluation was carried out through the utilization of a NaI(Tl) detector coupled with a PC-based multichannel analyzer. The acquired data were meticulously compared with established theoretical value. The results of this study point to a viable route for using this simple, cost-effective, and low-temperature synthesis approach to create nanomaterials suited for gamma ray shielding applications, as well as broader radiation protection. This novel technique has the potential to significantly improve radiation shielding technology. Along with this fast neutron attenuation capability of this prepared ferrite samples have been studied in terms of fast neutron removal cross section. • Ni–Cu–Zn Fe2O4 ferrite nanoparticles were synthesized utilizing the Co-precipitation technique. • Material preparation was confirmed using XRD, SEM, and FTIR. • The gamma ray and neutron shielding properties for ferrite materials were studied. • The experimental values have been supported with XCOM results. • Measured data are useful in nuclear medicine application. [ABSTRACT FROM AUTHOR]

The EU DEMO fusion reactor, which is part of the EUROfusion PPPT program and the successor to ITER, aims to facilitate the transition of fusion from the science laboratory to industrial use. Its key objective is net energy production. Its critical component, the divertor, manages power exhaust, ash, and helium removal. The key elements of the divertor include Plasma Facing Components (PFCs), shielding liners, reflector plates, divertor cassettes, and a vacuum vessel locking system. It is imperative that the shutdown dose rates are assessed to ensure workers' radiation protection during maintenance, plan the disposal of radioactive material, and ensure the overall safety of the plant. The shutdown dose rate tool employed in the present study is R2Smesh, which couples MCNP for radiation transport and FISPACT for neutron activation to calculate neutron-induced shutdown dose rates, utilizing an advanced, semi-heterogeneous EU DEMO divertor model, presenting a challenging task in neutron transport calculations. • R2Smesh was utilized for the evaluation of SDDR. • The analysis focused on SDDR values within a specific layer in the divertor. • 10 years after the shutdown, SDDR exhibits a significant reduction in Eurofer and Tungsten. • SDDR exhibits a significant reduction in EUROFER and Tungsten 10 years after the shutdown. • A four-order of magnitude drop in SDDR was noted in wishbone pins during a 100-year period. [ABSTRACT FROM AUTHOR]

Cyclotrons for the production of radiopharmaceuticals have become important tools in modern nuclear medicine. At the end of their lifecycles, such installations have to be dismantled and any activated materials must be treated according to the local radiation protection legislation. Using a simulation model, we have developed a non-destructive approach for the radiological characterization of components inside and around an IBA Cyclone 18/9 cyclotron. The methodology is based on software tools developed at CERN (FLUKA Monte Carlo code, ActiWiz and RAW). The simulation results were compared to measurements made using reference samples placed around the cyclotron inside the bunker. Results show a reasonable agreement between simulation and measurements of about a factor of two for a set of 27 reference samples and 11 radionuclides of interest. The origin of this factor has been thoroughly evaluated and opened the door to further investigations leading to possible avenues for improvement. • FLUKA modelling of an IBA cyclone 18/9 cyclotron including the liquid and gas targets. • Efficient method to calculate the radiological content inside materials using the software tools RAW and ActiWiz. • Experimental validation of the simulation results with the help of irradiated samples. • Determination of the simulation uncertainties thoroughly and in detail. [ABSTRACT FROM AUTHOR]

The borotellurite glasses whose chemical structure is (29.5–0.4x)CaO + 10CaF 2 + (60–0.6x)B 2 O 3 + xTeO 2 + 0.5Yb 2 O 3 (where x=10, 16, 22, 31, and 54 % mole. represent TCCBY1-TCCBY5, respectively) are Pb-free, thermally stable, and transparent glasses with attractive optical features for technological applications. The gamma-photons, electrons, protons, neutrons, carbon ions, fast neutrons, and fast neutron interaction parameters of these glasses are presented in this study to better understand the role of TeO 2 in influencing their radiation shielding properties and radiation protection applications. The photon mass attenuation coefficient was evaluated by XCOM computation and simulation using the FLUKA code. The FLUKA code was also used to evaluate the mass stopping powers of the charged radiations, while neutrons' cross sections were evaluated using standard expressions. For 0.015 MeV–15 MeV photons, the mass attenuation coefficients of the glasses fell from 17.9499 to 0.0246 cm2/g for TCCBY1, 20.5628 to 0.0263 cm2/g for TCCBY2, 23.2756 to 0.079 cm2/g for TCCBY3, 26.7487 to 0.0298 cm2/g for TCCBY4, and 33.3591 to 0.0335 cm2/g for TCCBY5. The photon half-value layer at 15 keV is reduced by about 19.57%, 32.68%, 48.84%, and 63.89% when the TeO 2 content increases from 10 mol to 16, 22, 31, and 54 mol, respectively. TeO 2 was found to suppress photon buildup in the glasses. The mass stopping powers of charged radiation increased as glass density decreased. The addition of TeO 2 into the glass structure increased the ability of the TCCBY glass to absorb fast neutrons by up to 54 % mole. The gamma radiation and fast neutron moderating ability of TCCBY5 glass compared to common shields and other materials is exceptional. The glass is recommended for the design of Pb-free, transparent, and efficient radiation protection structures. • Radiation shielding capacity of borotellurite glasses is estimated. • Monte Carlo simulations are carried out. • HVL is reduced by 63.89% at TeO 2 = 54 mol. • TCCBY5 glass system has a potential use in nuclear applications. [ABSTRACT FROM AUTHOR]

Compact Proton Therapy Centers, CPTC, have a single treatment room, and are technologically more affordable, smaller, advanced and easier to use. From a radiological protection point of view, the leading concern in CPTC are interactions of protons with components of the facility and patients that yield a broad emission of secondary particles, mainly high-energy neutrons, up to 230 MeV, and photons. Optimal design of shielding involves theoretical assumptions in the design phase and, consequently, experimental measurements with extended range neutron detectors must be carried out in the facility during the commissioning period to verify the design, assumptions and building of the enclosures. There are almost 50 CPTC under construction and planning around the world, hence the improvement of methodologies to verify the shielding and to evaluate the dose to workers and general public in CPTC is a trending issue. The aim of this work was to evaluate and compare the response of two commercial extended range REM meters, WENDI-II and LUPIN-II, for their application in shielding verification and radiation area monitoring in CPTC facilities, by estimating the ambient dose equivalent, H*(10), through the Monte Carlo code MCNP6. The results have been compared with previous works. Likewise, the performance evaluation of these devices in continuous energy neutron field have been carried out, using the AmBe/241 neutron source of the Neutronics Hall (NH) of the Neutron Measurements Laboratory of the Energy Engineering Department of Universidad Politecnica de Madrid (LMN-UPM), through Monte Carlo simulation with the MCNP6 code and experimental measurements. The work is framed into the project Contributions to Shielding and Dosimetry of Neutrons in CPTC. • CPTC involve advanced technologies and smaller footprint and represents the next step in the evolution of proton therapy. • Response of WENDI-II and LUPIN-II was assessed for its application in verification of shielding against neutrons in CPTC. • The benchmark in monoenergetic neutron fields was carried out calculating the ambient dose equivalent H*(10) with MCNP6. • The performance in continuous neutron fields have been carried out with experimental measurements using an AmBe/241 source. [ABSTRACT FROM AUTHOR]

For radiation protection optimization, ICRP proposed dose constraint as quantitative value for planned exposure situation based on representative person concept. The objective of this study is to derive dose constraints for the general public around nuclear power plants in Korea by applying representative person concept. The dose constraints for the general public around NPPs were derived through a total of six steps. The steps consisted of setting source terms, setting exposure pathways and scenarios, setting candidate groups for a critical group decision, setting habit data, calculating radiation doses, and proposing dose constraints. Through these steps, the radiation dose distribution of the general public around the NPPs was obtained, and dose constraints were proposed using the dose distribution. Radiation doses to the general public around all the Korea NPP sites ranged 1.63 × 10−2 to 1.32 × 10−1 mSv/y. Using the dose distribution, 0.15 mSv/y, 0.10 mSv/y, and 0.08 mSv/y were proposed as dose constraints. The dose constraint values derived in this study are proposals. Therefore, it is judged that the dose constraints should need furthermore discussion with regulators, licensees, and radiation protection experts considering societal and economic factors for radiation protection. The proposal for dose constraints developed in this study can be used to optimize radiation protection for the general public around the NPPs. • This study aims to derive the dose constraints for the general public around nuclear power plants (NPPs) in Korea for optimization of radiation protection • The dose constraints were derived using 10-year data of radioactive effluents released from all Korea NPPs and ICRP representative person concept. • Distribution of radiation doses to the general public around all the NPP sites in Korea ranged 1.63 × 10−2 to 1.32 × 10−1 mSv/y. • Three dose constraints, including 0.15 mSv/y, 0.10 mSv/y, and 0.08 mSv/y were finally proposed in this study. [ABSTRACT FROM AUTHOR]

During production of radiopharmaceuticals, the radiation situation in cyclotron pit is an important parameter, which is being monitored to ensure fulfilment of the limits and conditions of safe operation. The neutron flux in the structural components of the accelerator is also an important parameter, because the secondary neutrons are responsible for activation of cyclotron structural components and may even affect structural changes in it. This paper aims to characterize the neutron field in inner positions of medical accelerator IBA 18/9 by activation detectors and by means of scintillation spectrometry. The backward angle measurement was realized also in special liquid water target (H 2 18 O) at U120M cyclotron to confirm the data obtained in IBA 18/9 cyclotron. [Display omitted] • Neutron spectrum in backward angle (almost 180°) was measured with scintillation spectroscopy. • Neutron spectrum inside the production cyclotron was characterized. • Neutron spectra around a smaller target in open geometry were measured and compared with previous measurement. [ABSTRACT FROM AUTHOR]

The aim of this study is to improve the characterization and modeling of the radiation field surrounding the Leksell Gamma Knife®-PerfexionTM. The improved characterization of the radiation field enables more accurate shielding calculations to be performed for the areas adjacent to the treatment room. With the aid of a high-purity germanium detector and a satellite dose rate meter, γ -ray spectra and ambient dose equivalent H*(10) data were acquired at various locations in the field of a Leksell Gamma Knife unit in a treatment room at Karolinska University Hospital, Sweden. These measurements were used to validate the results of the PEGASOS Monte Carlo simulation system with a PENELOPE kernel. The levels of the radiation that passes through the shielding of the machine (leakage radiation) are shown to be much lower than what is suggested by various bodies, e.g. the National Council on Radiation Protection and Measurements, to be used when calculating radiation shielding barriers. The results clearly indicate that Monte Carlo simulations may be used in structural shielding design calculations for γ rays from the Leksell Gamma Knife. • A HPGe detector has been used to study the radiation field around the Gamma Knife; • A phase space has been used to study the radiation in the field of the Gamma Knife; • A novel pile-up modeling has been used to compare MC simulations with measurements. [ABSTRACT FROM AUTHOR]

In head and neck radiation therapy, the presence of dental restorations can increase unwanted neutron dose to the patient. This study aimed at the measurement of secondary neutron production induced by irradiation of a healthy tooth, Amalgam, Ni-Cr alloy and Ceramco with a photon beam generated in the treatment head of a Siemens Primus linac at a voltage of 15 MV. The irradiation field amounted to 10 × 10 cm2. The measurements of thermal and fast-neutron equivalent doses were performed by means of CR-39 detectors positioned in various depths of a Perspex (polymethyl methacrylate) phantom as at open field as at presence of corresponding dental restorations. The general trend of thermal neutron as well as fast-neutron equivalent dose behind the denture samples reveals their reduction with increasing depth. The maximum values of thermal-neutron dose related to Amalgam, Ceramco and Ni-Cr alloy amount to 1.45 mSv/100 MU, 1.38 mSv/100 MU and 1.32 mSv/100 MU, whereas the corresponding maximum values of fast-neutron dose at the depth of 1.8 cm amount to 0.19 mSv/100 MU, 1.04 mSv/100 MU and 0.97 mSv/100 MU, respectively. The present study investigates the neutron dose accompanied with radiotherapy. It is recommended that attempts have to be made to ensure that dental restorations are not in the path of the primary high-energy photon beam. Considering treatment planning, the guidelines of radiation protection should be improved. • Presence of dental restorations in path of photon beam in radiotherapy of head and neck cancer. • Amalgam, Ni-Cr alloy, and Ceramco are the most common dental restoration in cancerous patients. • Thermal and fast neutron production in radiotherapy with high energy photon due to their high density. • Amalgam, Ceramco, and Ni-Cr alloy have the dominant contribution in the total neutron dose, respectively. [ABSTRACT FROM AUTHOR]

The objective was to design a thin, flex card based personal dosemeter with low angular and energy dependence. It is based on silicon diodes that can measure the personal dose equivalent. Anisotropic conducting adhesive was used to connect the Si sensor to the flex card. Its intended use is for interventional radiology and after nuclear or radiological accidents, as it provides immediate detailed information about the dose rate to the wearer during shorter periods and integrates the dose rate during hours. It can be freely placed all over the body. By placing the dosemeter close to the eyes, it may be possible to estimate the personal dose equivalent as well as the dose at the lens of the eye. • A dosemeter based on sensitive Si detectors with low angular and energy dependence. • X-ray transparent mounting of Si-detector and flex card using ACA. • Helpful to the operator in clinical radiology and to first responder for radiation protection strategies. • Detailed information about dose rate and dose during shorter periods (0.3 ms to hours). • Energy filter design using Monte Carlo simulation and 3D metal printer. [ABSTRACT FROM AUTHOR]

Abstract Nowadays, high-energy X-rays produced by medical linear accelerators (LINACs) are widely used in many Radiation Therapy (RT) centers. High-energy photons (> 8 MeV) produce undesired neutrons in the LINAC head which raise concerns about unwanted neutron dose to the patients and RT personnel. Regarding the significance of radiation protection in RT, it is important to evaluate photoneutron contamination inside the RT room. Unfortunately, neutron dosimeters used for this purpose have limitations that can under the best conditions cause to > 10% uncertainty. In addition to this uncertainty, the present Monte Carlo (MC) study introduces another uncertainty in measurements (nearly up to 20%) when neutron ambient dose equivalent (H n * (10)) is measured at the patient table or inside the maze and the change in neutron energy is ignored. This type of uncertainty can even reach 35% if H n * (10) is measured by dosimeters covered by a layer of 10B as converter. So, in these cases, neglecting the change in neutron energy can threaten the credibility of measured data and one should attend to this energy change in order to reduce measurement uncertainty to the possible minimum. This study also discusses the change in neutron spectra and H n * (10) at the patient table caused by removing a typical RT room from MC simulations. Under such conditions, neutron mean energy ( Ē n ) overestimated by 0.2–0.4 MeV at the patient table. Neutron fluence ( φ n ) at the isocenter (IC) was underestimated by 23–54% for different field sizes that caused H n * (10) to be miscalculated up to 24%. This finding informs researchers that for accurate evaluation of H n * (10) at the patient table, simulating the RT room is an effective parameter in MC studies. Highlights • The MCNPX® code was used to simulate the photoneutron production in the head of a medical LINAC (Varian Clinac 2100C/D 18 MV). • Neutron ambient dose equivalent and energy spectrum were calculated for different field sizes at the patient table and inside the maze. • Neutron energy changed with field size and also detector location. • Neglecting the change in neutron energy lead to major uncertainty in dosimetry. • Neutron ambient dose equivalent at the patient table changed with removing Radiation Therapy room from simulations. [ABSTRACT FROM AUTHOR]

Abstract A Proton A ccelerator based B oron N eutron C apture T herapy (A-BNCT) facility is under development in Korea. Neutron beams for treatment are produced from a beryllium (Be) target and an 8 mA, 10 MeV proton beam. The purpose of the research is a radiation shielding analysis and an activation analysis for the facility design satisfying the radiation safety requirements as well as obtaining an operating license for the radiation facility according to a domestic nuclear commissioning procedure. The radiation shielding analysis was performed using the MCNPX computational particle transport code. The radiation source terms in the facility were evaluated and utilized in the shielding calculations. The minimum concrete thickness satisfying the designated dose rate of 5 μSv/h for the worker's area and 0.25 μSv/h for the public area were estimated and applied to the design. For an assessment of the radiation safety inside the facility, the dose rates were evaluated at several positions, such as behind the shielding door, around the primary barriers near the radiation sources, and in the penetrations of the ducts. The dose rate distribution was mapped for verification of the radiation safety for the entire facility. An activation analysis was carried out for the concrete walls, air, target assembly, beryllium target, and cooling water using FISPACT-2010 code. Concentrations of the activation products and dose rate induced by the radionuclides after shutdown were evaluated for the purpose of safe operation of the facility. The results were reviewed with the radiation safety regulations in Korea. As a result, it was proved that the final facility design satisfies the safety requirements. Highlights • A Proton A ccelerator based B oron N eutron C apture T herapy (A-BNCT) facility is under development in Korea. • Radiation shielding analysis was carried out using MCNPX code • The facility design was determined satisfying designated dose limit • Neutron activation analysis was performed with MCNPX and FISPACT code • The concentration of radionuclides were reviewed with the radiation safety regulations. [ABSTRACT FROM AUTHOR]

Abstract In this study, clay brick (CBRICK) and pumice brick (PBRICK) samples used as structural material in the construction of dwellings, schools, workplaces and factories in Turkey were compared with each other from a radiological viewpoint. The activity concentrations of 226Ra, 232Th and 40K naturally occurring in CBRICK and PBRICK samples collected from different regions of Turkey were determined by using a gamma-ray spectrometer with a high purity germanium detector (HPGe). The average activity concentrations of 226Ra, 232Th and 40K in CBRICK and PBRICK samples were found as 35.4 ± 3.3, 37.5 ± 2.4 and 593.7 ± 42.7 Bq kg−1 and 81.9 ± 2.5, 65.8 ± 6.0 and 1066.0 ± 46.6 Bq kg−1, respectively. The radon surface exhalation rate (EX S) and radon mass exhalation rate (EX M) of CBRICK and PBRICK samples were measured by using an active radon gas analyzer with an accumulation container. The average value of EX S and EX M of CBRICK and PBRICK samples found as 45.9 ± 2.9 mBq m−2 h−1 and 3.7 ± 2.9 mBq kg−1 h−1 and 100.9 ± 4.7 mBq m−2 h−1 and 9.9 ± 0.5 mBq kg−1 h−1, respectively. Radiological parameters related to external and internal exposure to members of the public such as the radiation protection index, alpha index, and indoor absorbed gamma radiation dose rate and the corresponding annual effective dose from external exposure, annual effective dose from inhalation of radon, and the lifetime cancer risk were estimated for CBRICK and PBRICK samples. The results were compared with each other and with the international recommended limits or criteria. The results reveal that the average values measured and estimated for CBRICK samples are approximately two times lower than those measured and estimated for PBRICK samples. Thus, from the radiological viewpoint, clay brick is preferable to pumice brick as a structural material in the building sector. Highlights • Radiological comparison of the use of clay and pumice bricks was made. • The radioactivity in clay and pumice bricks was determined using a gamma-ray spectrometry. • Radon exhalation rates from brick samples was firstly measured using a radon monitor system. • The radiation protection index and alpha index were estimated for each brick sample. • Externally absorbed indoor gamma dose rate and annual effective dose were estimated. [ABSTRACT FROM AUTHOR]

The aim of this study is to develop the Effective Dose Calculation Program (EDCP) for the usage of Naturally Occurring Radioactive Material (NORM) added consumer products. The EDCP was developed based on a database of effective dose conversion coefficient and the Matrix Laboratory (MATLAB) program to incorporate a Graphic User Interface (GUI) for ease of use. To validate EDCP, the effective dose calculated with EDCP by manually determining the source region by using the GUI and that by using the reference mathematical algorithm were compared for pillow, waist supporter, eye-patch and sleeping mattress. The results show that the annual effective dose calculated with EDCP was almost identical to that calculated using the reference mathematical algorithm in most of the assessment cases. With the assumption of the gamma energy of 1 MeV and activity of 1 MBq, the annual effective doses of pillow, waist supporter, sleeping mattress, and eye-patch determined using the reference algorithm were 3.444 mSv year −1 , 2.770 mSv year −1 , 4.629 mSv year −1 , and 3.567 mSv year −1 , respectively, while those calculated using EDCP were 3.561 mSv year −1 , 2.630 mSv year −1 , 4.740 mSv year −1 , and 3.780 mSv year −1 , respectively. The differences in the annual effective doses were less than 5%, despite the different calculation methods employed. The EDCP can therefore be effectively used for radiation protection management in the context of the usage of NORM-added consumer products. Additionally, EDCP can be used by members of the public through the GUI for various studies in the field of radiation protection, thus facilitating easy access to the program. [ABSTRACT FROM AUTHOR]

High-density polyethylene (HDPE) was obtained through a compression molding technique, and incorporated with different filler weight fractions (0, 10, 15, 25, and 35%) of bulk WO 3 , and two different WO 3 nanoparticle sizes (45 nm and 24 nm). The radiation attenuation ability of the new category of polymer composite HDPE/WO 3 was evaluated using gamma-ray energies ranging from 59.53 up to 1332.5keV of four radioactive sources 241Am, 133Ba, 137Cs, and 60Co. The mass attenuation coefficients μ m , the total molecular cross-section σ m o l , the effective atomic cross-section σ a t o m , the total electronic cross-section σ e l , the effective atomic number Z e f f , electron density N e f f , the half value layer (HVL), the tenth value layer (TVL), and the relaxation length were investigated. The obtained results of the gamma-ray attenuation parameters exhibited an outstanding influence of the size and weight fraction of WO 3 filler on the gamma-ray shielding ability of the HDPE composite. A significant improvement was detected at low gamma-ray energies. The HVL of the synthesized HDPE composites is compared with that of pure lead as a conventional shielding material. HDPE composite filled with the smaller size of WO 3 nanoparticle shows good improvement in the attenuation parameters, which suggests promising applications in radiation protection and gamma-ray shielding. • Investigated the effect of particle size and weight fraction of WO 3 filler on the gamma-ray attenuation ability. • Attenuation parameters exhibit significant improvement by increasing the weight fraction of WO 3. • Nanoparticle size of WO 3 shows a better ability to attenuate lower gamma-ray energies. [ABSTRACT FROM AUTHOR]

In this research paper, we report on the radiation shielding ability of five newly developed chalcogenide alloys described by the chemical compositions of Ge 20 Sb 6 Te 72 Bi 2 (GTSB1), Ge 20 Sb 6 Te 70 Bi 4 (GTSB2), Ge 20 Sb 6 Te 68 Bi 6 (GTSB3), Ge 20 Sb 6 Te 66 Bi 8 (GTSB4), and Ge 20 Sb 6 Te 64 Bi 10 (GTSB5). The simulation (Monte Carlo) technique is systematically utilized to understand the problem of radiation propagation through the chalcogenide alloys. The maximum deviation between the theoretical values and simulations outcomes for each alloy sample corresponds to circa 0.525, 0.517, 0.875, 0.619, and 0.574% for GTSB1, GTSB2, GTSB3, GTSB4, and GTSB5, respectively. The obtained results indicate that the rapid decline in the value of the attenuation coefficients is mainly due to the main photon interaction process with the alloys for E ≤ 500 keV. Additionally, the charged particle and neutron transmission properties are assessed for the involved chalcogenide alloys. Comparing the MFP and HVL with those of the conventional shielding glasses and concretes, the present alloys are good photon absorbers and they could be used to replace some conventional shields in radiation protection applications. • High Te content chalcogenide alloys were reported for radiation shielding applications. • Monte Carlo method was successfully applied in the design of vacuum system for radiation propagation. • The optimum thickness for narrow beam transmission geometry is less than 0.1 cm. • GTSB5 alloys can conveniently replace the traditional materials. [ABSTRACT FROM AUTHOR]

In the framework of the EMPIR project traceRadon, stable atmospheres with low-level radon activity concentrations have to be produced for calibrating radon detectors designed to measure outdoor air activity concentrations. The traceable calibration of these detectors at very low activity concentrations is of special interest to the radiation protection, climate observation, and atmospheric research communities. Radiation protection networks (such as the EUropean Radiological Data Exchange Platform (EURDEP)) and atmospheric monitoring networks (such as the Integrated Carbon Observation System (ICOS)) need reliable and accurate radon activity concentration measurements for a variety of reasons, including: the identification of Radon Priority Areas (RPA); improving the sensitivity and reliability of radiological emergency early warning systems (Melintescu et al., 2018); for more reliable application of the Radon Tracer Method (RTM) to estimate greenhouse gas (GHG) emissions; for improved global "baseline" monitoring of changing GHG concentrations and quantification of regional pollution transport (Chambers et al., 2016), (Chambers et al., 2018); and for evaluating mixing and transport parameterisations in regional or global chemical transport models (CTMs) (Zhang et al., 2021), (Chambers et al., 2019). To achieve this goal, low activity sources of radium with a variety of characteristics were produced using different methods. Sources ranging from MBq 226Ra down to several Bq 226Ra were developed and characterised during the evolution of production methods, and uncertainties below 2 % (k=1) were achieved through dedicated detection techniques, even for the lowest activity sources. The uncertainty of the lowest activity sources was improved using a new online measurement technique for which the source and detector were combined in the same device. This Integrated Radon Source Detector device, henceforth an IRSD, reaches a counting efficiency approaching 50 % through detection under quasi 2π sr solid-angle. At the time of this study the IRSD was already produced with 226Ra activities between 2 Bq and 440 Bq. To compare the working performance of the developed sources (i.e., to establish a reference atmosphere), study the stability of the sources, and to establish traceability to national standards, an intercomparison exercise was carried out at the PTB facility. Here we present the various source production techniques, the determination of their radium activity, and determination of their radon emanation (including assigned uncertainties). This includes details of the implementation of the intercomparison set-up, and a discussion of the results of the source characterisations. [ABSTRACT FROM AUTHOR]

After a nuclear or radiation event, emergency responders and radiation protection authorities need quick and credible information based on reliable accident and post-accident radiological data. However, risks to people in the vicinity of the source pose serious measurement challenges. Many problems could be solved by unmanned airborne monitoring systems, but the current ones are mostly based on non-spectrometric detectors carried by drones with low bearing, short flying range and flight time. Therefore spectrometric monitoring system based on High-Purity Germanium (HPGe) detector carried by powerful unmanned helicopter has been developed. The presented unmanned aerial spectrometric system is reliable and heavy-duty and enables quick and safe identification of released radionuclides, thus provides a basis for determining the plant damage state and for planning of emergency and contamination zones. The system will support timely, effective actions that protect the public and environment against the effects of ionizing radiation. The paper describes development and performance tests of this novel system. • Unique airborne HPGe spectrometric system developed. • Unmanned helicopter with high payload capacity and HPGe detector used. • Emergency responders will be provided with list of released radionuclides. • Performance tested using standardized sources and Monte Carlo simulations. [ABSTRACT FROM AUTHOR]

Precise radon measurements are a requirement for various applications, ranging from radiation protection over environmental studies to material screening campaigns for rare-event searches. All of them ultimately depend on the availability of calibration sources with a known and stable radon emanation rate. A new approach to produce clean and dry radon sources by implantation of 226Ra ions into stainless steel has been investigated. In a proof of principle study, two stainless steel plates have been implanted in collaboration with the ISOLDE facility located at CERN. We present results from a complete characterization of the sources. Each sample provides a radon emanation rate of about 2 Bq, which has been measured using electrostatic radon monitors as well as miniaturized proportional counters. Additional measurements using HPGe and alpha spectrometry as well as measurements of the radon emanation rate at low temperatures were carried out. • Production of 222Rn emanating stainless steel sources by implantation of 226Ra. • Measurement of the implanted activity (8 Bq) with HPGe- and alpha-spectrometry. • Radon emanation rate (2 Bq) measured by radon monitor and proportional counter. • Recoil dominated emanation without significant dependence on temperature and pressure. [ABSTRACT FROM AUTHOR]

In this paper, a comprehensive overview on the achievements and generated research results beyond the state-of-the-art is given based on the working structure of the joint metrology research project MetroRADON. The results of the project have been targeted at the implementation of the European Council Directive 2013/59/EURATOM on radiation protection (EU BSS) and benefit European and international standards on radon monitoring, radon measurement and calibration, geographical radon mapping, and guidelines on radiological protection, construction products, radiation instrumentation and nuclear data. • Traceable calibration of radon (222Rn) instruments at low activity concentrations. • Reducing the influence of thoron, temperature, and humidity on radon measurements. • Comprehensive study on existing radon measurement procedures in Europe. • "Geogenic Radon Hazard Index" (GRHI) - development of a new tool for radon mapping. • Calibration and measurement procedures for radon concentration in air. [ABSTRACT FROM AUTHOR]

Effective dose (E) in medical procedures is of practical value for comparing doses from different types of procedures which is not possible with the different measurable dose quantities. In this survey we estimated and compared the values of E in 12 medical imaging procedures. The mean E for conventional X-ray in mSv were ranged between 0.01 for skull lateral to 0.5 for abdominal while the mean E for cardiac interventional radiology in mSv was up to 16 for percutaneous transluminal coronary angioplasty (PTCA). For dental radiology the mean E in μSv was 1.2 for intraoral and 10 for panoramic. In mammography the mean E was 0.5 mSv. Compared with the literature, chest postero–anterior (PA), lumbar spine AP, PTCA and mammography procedures had shown larger E values. The obtained results can help in justification of techniques. [ABSTRACT FROM AUTHOR]

In the present work the effect of tungsten oxide (WO 3 ) nanoparticles on mass attenauation coefficients of concrete has been investigated by using MCNPX (version 2.4.0). The validation of generated MCNPX simulation geometry has been provided by comparing the results with standard XCOM data for mass attenuation coefficients of concrete. A very good agreement between XCOM and MCNPX have been obtained. The validated geometry has been used for definition of nano-WO 3 and micro-WO 3 into concrete sample. The mass attenuation coefficients of pure concrete and WO 3 added concrete with micro-sized and nano-sized have been compared. It was observed that shielding properties of concrete doped with WO 3 increased. The results of mass attenauation coefficients also showed that the concrete doped with nano-WO 3 significanlty improve shielding properties than micro-WO 3 . It can be concluded that addition of nano-sized particles can be considered as another mechanism to reduce radiation dose. [ABSTRACT FROM AUTHOR]

Delayed neutron counting is often used to verify the characteristics of nuclear material. Use of portable neutron generators in high frequency pulsing mode enables effective analysis with higher counting efficiency and lower radiation protection demands. The paper deals with delayed neutron counting of uranium pins with P385 portable DD neutron generator in polyethylene-based setup. Counting is performed in high frequency mode of neutron generator. The viability of such measurement in the frequency range from 100 Hz to 250 Hz was demonstrated and optimal pulsing parameters for P385 neutron generator were found. Delayed neutron counting was then performed for two types of uranium rods. Delayed neutrons were counted both inbetween neutron pulses during neutron generation and once the emission of neutrons stopped. The results were further validated by Monte Carlo (MCNP) simulations. For the geometry of studied rods, the MCNP calculation were done to calculate the dependence of the response to delayed neutrons on rod enrichment, to show the viability to use the method for rod enrichment verification. An option of using cadmium inset in irradiation channel to overcome the effect of self-shielding for samples with higher enrichment was proposed, experimentally tested, and evaluated through MCNP calculations. • Portable DD neutron generator was used for delayed neutron counting. • DD generator and data acquisition setup was optimized. • Applicability for enrichment verification of fuel pins was tested. • Study includes interrogation of uranium pins by thermal and epithermal neutrons. • Monte Carlo simulation was used to handle effects of self-shielding, geometry, and composition. [ABSTRACT FROM AUTHOR]

Beta Secondary Standard 2 (BSS 2) provides beta radiation fields with certified values of absorbed dose to tissue and the derived operational radiation protection quantities. As part of the quality assurance, the reliability of the CDTN BSS2 system was verified through measurements in the 90 Sr/ 90 Y and 85 Kr beta radiation fields. Absorbed dose rates and their angular variation were measured with a 23392 model PTW extrapolation chamber and with Gafchromic radiochromic films on a PMMA slab phantom. The feasibility of using both methods was analyzed. [ABSTRACT FROM AUTHOR]

A Bonner sphere spectrometer was used to measure the neutron spectra produced at the collision of protons with an H 2 18 O target at different angles. A unique H 2 18 O target to produce 18 F was designed and placed in a Tandem linear particle accelerator which produces 8.5 MeV protons. The neutron count rates measured with the Bonner spheres were unfolded with the MAXED code. With the GEANT4 Monte Carlo code the neutron spectrum induced in the (p, n) reaction was estimated, this spectrum was used as initial guess during unfolding. Although the cross section of the reaction 18 O(p,n) 18 F is well known, the neutron energy spectra is not correctly defined and it is necessary to verify the simulation with measurements. For this reason, the sensitivity of the unfolding method to the initial spectrum was analyzed applying small variation to the fast neutron peak. [ABSTRACT FROM AUTHOR]

Proton therapy is becoming increasingly popular worldwide, and its shielding must be considered. The cathode ray tube (CRT) material is a glass containing heavy metal elements, these materials have become a good choice for the production of radiation-proof concrete. In this study, the ability of concrete containing CRT fragments as shielding materials for proton therapy rooms is evaluated in terms of neutron shielding ability, neutron reflection ability, ambient dose equivalent rate, and induced radioactivity. In addition, this concrete is compared with commonly used ordinary concrete, boron-containing concrete, and barite concrete. The results show that with the increase of CRT content (10%–90%), the transmitted neutron fluence decreases continuously (5.06 × 10−10 – 1.77 × 10−10 cm−2/particle), and the reflection of neutrons gradually increases (2.64 × 10−9 – 3.20 × 10−9 cm−2/particle), resulting in an increased potential to patients. When 50% CRT concrete is used, the ambient dose equivalent rate is below 3.80 μSv/h/nA, and 90% CRT concrete is below 3.11 μSv/h/nA. The trend of radionuclide activity of induced radioactivity from 0 to 60 min after irradiation for concrete with different CRT contents is 2.74–5.38 × 10−3 Bq/cm3, and the maximum photon fluence is 8.13 × 102 cm−2. In conclusion, the optimization model of the three-layer shielding structure of ordinary concrete, high CRT content concrete, and boron-containing concrete is proposed with ambient dose equivalent rate less than 1.88 μSv/h/nA, minimizing the reflected neutrons to which the patient is exposed. This study shows the protection performance of CRT concrete is better than ordinary concrete and barite concrete. • The ability of CRT concrete as a shielding materials for proton therapy is evaluated and compared. • Neutron shielding and reflection ability, ambient dose equivalent rate, induced radioactivity are considered. • The optimization model of the three-layer shielding structure is proposed. [ABSTRACT FROM AUTHOR]

In the current study, we fabricated a series of boro-tellurite glass samples with a composition 10SrO–10Al 2 O 3 –10MoO 3 -(70-x)B 2 O 3 -xTeO 2 , where (x = 0, 17.5, 35, 52.5, and 70 mol%) via an ordinary melt-quench method. The glass structure was explored by X-ray diffraction (XRD), physical and structural properties. XRD results affirm the existence of two broad peaks, proving the amorphous state of the current glasses. The acquired results exhibit a linear relationship between the density, Poisson's ratio, and the addition of TeO 2 amounts. The addition of TeO 2 to the glass system shows a rise in glass stability and a reduction in packing density. Additionally, the values of mass attenuation coefficient (MAC) were determined experimentally within five energies (0.184, 0.280, 0.661, 0.710, and 0.810 MeV) from two radiation sources (166Ho and 137Cs). The (MAC) exp results were compared with XCOM values, and the compared values showed excellent compatibility. From the experimental results, many radiation shielding features involving half-value layer (HVL), mean free path (MFP), tenth value layer (TVL), and radiation protection efficiency (RPE) were computed. From the obtained results, it can be concluded that the TeSB4 sample has the highest stability and absorption for radiation, indicating the ability to use it as a radiation shielding substance. • A new series of glass samples were prepared by ordinary melt-quench method. • The prepared samples appeared attractive results in the radiation shielding field. • The shielding properties of all prepared samples were measured experimentally. [ABSTRACT FROM AUTHOR]

Intakes of radionuclides can be estimated through in vivo measurements, and the uncertainties associated to the measured activities should be clearly stated in monitoring program reports. This study aims to evaluate the uncertainties of in vivo monitoring of iodine 131 in the thyroid. The reference values for high-energy photons are based on the IDEAS Guide. Measurements were performed at the In Vivo Monitoring Laboratory of the Institute of Radiation Protection and Dosimetry (IRD) and at the Internal Dosimetry Laboratory of the Regional Center of Nuclear Sciences (CRCN-NE). In both institutions, the experiment was performed using a NaI(Tl) 3′′3″ scintillation detector and a neck-thyroid phantom. Scattering factors were calculated and compared in different counting geometries. The results show that the technique produces reproducibility equivalent to the values suggested in the IDEAS Guide and measurement uncertainties is comparable to international quality standards for this type of in vivo monitoring. [ABSTRACT FROM AUTHOR]

Zirconia was investigated as a radiation-resistant support for radium and actinium aqueous ion separations. Acetate and total metal ion concentrations were experimentally evaluated as factors that could affect retention using La and Ba surrogates. Pseudo-second order rate constants were derived and the uptake of La was determined to be endothermic. Elution profiles containing Ba, La, 228Ra, 228Ac, and 212Pb are presented and discussed. • Actinium-225 is being developed for targeted alpha therapy and can be produced using radium targets. • Actinium and radium can induce high alpha doses to organic resins that complicate separations. • Inorganic supports such as zirconia provide radiation protection and efficient metal ion separations. • Specific pH windows and background buffer concentrations were determined to optimize separations of isotopes. [ABSTRACT FROM AUTHOR]

A computational simulation of alpha-particle expansion, emitted by 222Rn naturally diluted in the water of a spring located in the state of Puebla Mexico, is presented. This simulation provided information on the volume of expansion of the alpha particles in the vicinity of the spring and thus awareness if there was a radiological risk for the users or the population that lived near to it. Before performing the simulation, several measurements were made to water samples with a dynamic measurement system. This in order to know the level of radon concentration and compare the results obtained with the levels recommended by the United States Environmental Protection Agency (US-EPA) and the World Health Organization (WHO). In addition, to know if there was a radiological risk due to the presence of gamma emitting radionuclides, complementary water analyses were carried out using gamma-ray spectrometry techniques. The simulation was developed using the scientific software of particle interaction with matter, Geant4. The different variables declared for the software parameters are presented in this document. The results of the radon measurements in the water and the computational simulation, determined that there was no radiological risk due to alpha radiation. Furthermore, the results from the gamma-ray spectrometer showed that there was no presence of other hazardous radionuclides in the water. • Computer simulations. • Radon in drinking water. • Environmental pollution. [ABSTRACT FROM AUTHOR]

In 2020, the International Commission on Radiation Units & Measurements ICRU has released Report 95 on "Operational Quantities for External Radiation Exposure". This publication introduced a new measurand, namely ambient dose H ∗ , as the operational quantity for external exposure to be applied in the future replacing ambient dose equivalent H ∗ (10). It should be noted that this change will make it necessary to adjust previously used constants and coefficients or at least to review them with regard to the new measurand. Therefore, this recommendation represents a significant cut in implementation of radiation protection systems. The revision of all previously valid factors makes it possible to develop a consistent data set. The combination of experience gained in the last decades with more extensive data sets (e.g. on the spectra of individual radionuclides) with comparatively large computing power allows the definition of optimized data collections serving as input for Monte Carlo methods. Therefore, this publication is intended to lay a suitable foundation consisting of dose rate constants for more than 400 radionuclides concerning the future measurand ambient dose H ∗ and to indicate differences to values related to the ambient dose equivalent H ∗ (10) used nowadays. [Display omitted] • Calculation of dose rate constants in ambient dose for more than 400 radionuclides. • Comparison of different conversion coefficients published by ISO 4037-3 as well as ICRU 95. • Comparison of numerical values of dose rate constants with discussion of differences. [ABSTRACT FROM AUTHOR]

This study intends to measure the radiation dose to patients and staff during (i) Dynamic Hip Screw (DHS) and (ii) Dynamic Cannula Screw (DCS) and to evaluate entrance surface Air kerma (ESAK) dose and organ doses and effective doses. Calibrated Thermoluminescence dosimeters (TLD-GR200A) were used. The mean patients’ doses were 0.46 mGy and 0.07 mGy for DHS and DCS procedures, respectively. The mean staff doses at the thyroid and chest were 4.69 mGy and 1.21 mGy per procedure. The mean organ and effective dose for patients and staff were higher in DHS compared to DCS. Orthopedic surgeons were exposed to unnecessary radiation doses due to the lack of protection measures. The radiation dose per hip procedure is within the safety limit and less than the previous studies. [ABSTRACT FROM AUTHOR]

This study aims to evaluate and compare protocols with similar purposes in a cone beam CT scanner using thermoluminescent dosimeter (TLD) and the air kerma-area product ( P KA ) as the kerma index. The measurements were performed on two protocols used to obtain an image of the maxilla–mandible using the equipment GENDEX GXCB 500: Protocol [GX1] extended diameter and asymmetric beam (14 cm×8.5 cm-maxilla/mandible) and protocol [GX2] symmetrical beam (8.5 cm×8.5 cm-maxillary/mandible). LiF dosimeters inserted into a female anthropomorphic phantom were used. For both protocols, the value of P KA was evaluated using a PTW Diamentor E2 meter and the multimeter Radcal Rapidose system. The results obtained for the effective dose/ P KA were separated by protocol image. [GX1]: 44.5 µSv/478 mGy cm 2 ; [GX2]: 54.8 µSv/507 mGy cm 2 . Although the ratio of the diameters (14 cm/8.5 cm)=1.65, the ratio of effective dose values (44.5 µSv/54.8 µSv)=0.81, that is, the effective dose of the protocol with extended diameter is 19% smaller. The P KA values reveal very similar results between the two protocols. For the cases where the scanner uses an asymmetric beam to obtain images with large diameters that cover the entire face, there are advantages from the point of view of reducing the exposure of patients when compared to the use of symmetrical beam and/or to FOV images with a smaller diameter. [ABSTRACT FROM AUTHOR]

Recent studies have shown that D-T neutron generator can be used as a proper neutron source for Boron Neutron Capture Therapy (BNCT) of deep-seated brain tumors. In this paper, radiation shielding calculations have been conducted based on the computational method for designing a BNCT treatment room for a recent proposed D-T neutron source. By using the MCNP-4C code, the geometry of the treatment room has been designed and optimized in such a way that the equivalent dose rate out of the treatment room to be less than 0.5 μSv/h for uncontrolled areas. The treatment room contains walls, monitoring window, maze and entrance door. According to the radiation protection viewpoint, dose rate results of out of the proposed room showed that using D-T neutron source for BNCT is safe. [ABSTRACT FROM AUTHOR]

CR-39 solid-state nuclear track detectors (SSNTDs) were used to study the variations of fast neutron relative intensities around a high-energy (23 MV) linear accelerator (Varian 21EX) photon beam. The variations were determined on the patient plane at 0, 50, 100, 150 and 200 cm from the isocenter of the photon beam. In addition, photoneutron intensities and distributions at isocenter level with field size of 40×40 cm 2 at Source Axis Distance (SAD)=100 cm around 23 MV photon beam were also determined. The results showed that the photoneutron intensities decreased rapidly by increasing the distance from the center of the x-ray beam towards the periphery, for the open fields. [ABSTRACT FROM AUTHOR]

Accurate determination of calibration factors for radionuclide activity meters is crucial for quantitative studies and in the optimization step of radiation protection, as these detectors are widespread in radiopharmacy and nuclear medicine facilities. In this work we developed the Monte Carlo model of a widely used activity meter, using the Geant4 simulation toolkit. More precisely the “PENELOPE” EM physics models were employed. The model was validated by means of several certified sources, traceable to primary activity standards, and other sources locally standardized with spectrometry measurements, plus other experimental tests. Great care was taken in order to accurately reproduce the geometrical details of the gas chamber and the activity sources, each of which is different in shape and enclosed in a unique container. Both relative calibration factors and ionization current obtained with simulations were compared against experimental measurements; further tests were carried out, such as the comparison of the relative response of the chamber for a source placed at different positions. The results showed a satisfactory level of accuracy in the energy range of interest, with the discrepancies lower than 4% for all the tested parameters. This shows that an accurate Monte Carlo modeling of this type of detector is feasible using the low-energy physics models embedded in Geant4. The obtained Monte Carlo model establishes a powerful tool for first instance determination of new calibration factors for non-standard radionuclides, for custom containers, when a reference source is not available. Moreover, the model provides an experimental setup for further research and optimization with regards to materials and geometrical details of the measuring setup, such as the ionization chamber itself or the containers configuration. [ABSTRACT FROM AUTHOR]

As a consequence of the Chernobyl accident in 1986 the Integrated Measurement and Information System (IMIS) was established (Weiss and Leeb, 1993) which includes on-line monitoring networks for the surveillance of radioactivity in Germany. Today, the German Federal Office for Radiation Protection (BfS) operates a gamma dose rate network with 1800 ambient dose equivalent rate H*(10) (ADER) stations almost equally distributed over the German territory. The ADER network integrates Geiger-Müller (GM) based detectors which, if low and high dose rate tubes are combined, are known to have excellent long-term stability and an extended dose rate range from environmental background level (20 nSv/h) up to several Sv/h. However, one main drawback is the lack of information about nuclides contributing to the observed dose rate. Therefore BfS has started to integrate LaBr 3 -based spectrometric detector systems (so-called spectro-dosemeters) in the existing ADER network. In this paper detector design, quality assurance and quality control (QA/QC) procedures are described as well as efforts required to characterize and operate monitoring networks based on spectrometric detectors. • Integration of spectro-dosemeters into the German ADER network. • Characterization and initial testing of new spectrometric detectors. • Development and application of the spectrum analysis software. [ABSTRACT FROM AUTHOR]

Staff occupational radiation exposure is limited to 20 mSv annually to preclude tissue reaction and lower risk of cancer effect. Staff occupational exposure arises during the preparation, injection, and scanning of the patients. Recent studies reported that nuclear medicine personnel might exceed the annual dose limit in high workload and poor radiation protection circumstances. Therefore, an accurate estimation of the annual dose limit is recommended. The goal of this research is to calculate the cumulative external effective dose (mSv) per year for nuclear medicine physicians, technologists, and nurses at SPECT/CT department. A total of 15 staff worked in the nuclear medicine department at King Saud Medical City (KSMC), Riyadh, Saudi Arabia were evaluated for the last six years. 99mTc is used more frequently for most of the patients. The procedures include renal, cardiac scintigraphy procedures. Staff dose was quantified using calibrated thermoluminecnt dosimeters (TLD-100) with an automatic TLD reader (Harshaw 6600). Exposure to ionizing radiation was evaluated in terms of deep doses (Hp(10) were evaluated. The overall average and standard deviation of the external doses for nuclear medicine physicians, technologists' and nurses were 1.8 ± 0.7, 1.9 ± 0.6, 2.0 ± 0.9, 2.2 ± 0.8, 6.0 ± 2.8, and 3.6 ± 1.3 for the years 2015,2016,2017,2018,2019, and 2020, respectively. Technologists and nurses received higher doses of compared to the nuclear medicine physicians. Technologists and nurses involved in radionuclide preparation, patients' injection, and image acquisition. Staff annual exposure is below the annual dose limits; however, this external dose is considered high compared to the current workload. • Staff annual effective dose was estimated at nuclear medicine department equipped with two SPECT/CT systems. • External effective dose in terms of Hp(10) was monitored for six consecutive years. • Occupational exposure was quantified with thermoluminecnt dosimeters (TLD-100). • Technologists and nurses received higher doses than the nuclear medicine physicians. • Staff annual radiation from external exposure is 50% below the annual dose limits. [ABSTRACT FROM AUTHOR]

Safe, effective, cost-effective, easy feasible and low-waste decontamination technologies are fundamental importance from environmental and radiation protection aspects. In this study the effectiveness of AP-CITROX decontamination technology of Inconel alloy 690 was investigated. Non-radioactive representative metal samples were formed to test of decontamination technology and the clear-, the corroded-, the decontaminated layer were analysed electrochemically. The results indicate that the passivation step of the technology was not completed. • Effectiveness of AP-CITROX decontamination technology of inconel alloy. • Oxide layer formation on inconel alloys. • Corrosion properties of inconel alloys. [ABSTRACT FROM AUTHOR]

Abstract: In this paper the assessment of natural radionuclides' activity concentration in building materials, calibration requirements and related indoor exposure dose models is presented. Particular attention is turned to specific improvements in low-level gamma-ray spectrometry to determine the activity concentration of necessary natural radionuclides in building materials with adequate measurement uncertainties. Different approaches for the modelling of the effective dose indoor due to external radiation resulted from natural radionuclides in building material and results of actual building material assessments are shown. [Copyright &y& Elsevier]

The out-diffusion of radionuclides from activated material in case of a fire may represent a non-negligible contribution to the radiological source term of such an event. In order to assess the contribution of this phenomenon, a software package has been designed and implemented. In the present document we briefly introduce the numerical treatment used to tackle the problem prior to the explanation of the software's logic. The document ends with an exemplary simulation and a study carried out to validate the implementation of the algorithm. The presented tool has been named SOLIDUSS, it is mainly written in C++ and uses a Monte Carlo based approach to simulate the diffusion of radioisotopes within solid materials. It is designed to run coupled with CERN-FLUKA, taking advantage of its geometry kernel to carry out diffusion calculations in arbitrarily complex geometries. The user can provide 3D temperature maps along with many other parameters that allow the program to target a wide range of different scenarios. As results SOLIDUSS provides 3D radionuclide concentration maps as well as the amount of radionuclides out-diffused from the selected materials. So far, this software has undergone numerical validation which will be discussed in this paper. Benchmarking against experimental data is currently ongoing. • Out-diffusion of radionuclides in case of fire may contribute to its source term. • A software named SOLIDUSS was developed to assess the out-diffusion of isotopes. • SOLIDUSS is Monte Carlo based and runs coupled with CERN-FLUKA. • Diffusion is simulated for arbitrarily complex geometries using a Wiener process. • Radioisotopes can be tracked according to 3-D temperature maps. [ABSTRACT FROM AUTHOR]

The nuclide-specific dose rate constant, formerly called gamma ray constant, is one of the most important quantities in practical radiation protection dosimetry. For radiation sources with known radionuclide composition and activity, the expected dose rates at various distances can easily be calculated with reasonable approximations. In addition, they serve as a planning basis for the design of shielding of radiation application rooms and facilities. In this study, dose rate constants were calculated using the most recent conversion coefficients and the most suitable spectral data for more than 400 radionuclides using different calculation approaches. In addition this paper provides a critical review of currently published dose rate constants for the ambient dose equivalent H ∗ (10). [Display omitted] • In a novel approach, x-rays are considered for dose rate constants and applied for the calculation of 400 radionuclides. • Comparison of different approaches for calculating dose rate constants. • Comparison of different conversion coefficients published bei IRCU 57, ICRP 74 and ISO 4037-3. • Evaluation of published dose rate constants for radionuclides in different energy ranges. • Investigation of possible causes for the discrepancies between the published data. [ABSTRACT FROM AUTHOR]

Purpose of this study is to investigate effect of gadolinium (Gd) content of GdxFe2O3(100-x), (x = 0, 5, 10, 20) composite system on structural, thermal, optical and radiation protection of poly (vinyl alchool) PVA based nanocomposite film. In this study, Gd doped Fe 2 O 3 nanopowders (with Gd = 0, 5, 10 and 20) were synthesized by spray pyrolysis method. The characterization of these powders have been obtained by using X-ray diffractometer (XRD), scanning electron microscopy (SEM), and Uv–vis spectrometer. After then, GdxFe 2 O 3 (100-x)/PVA (x = 0, 5, 10, 20) polymer composite systems were prepared by solution casting technique. The structural, thermal and optical properties of obtained polymer composites have been examined. The irradiation properties of 6 MeV energized X-ray via cLINAC to obtain radiation shielding performance of these different percentage (0, 5, 10, and 20%) Gd doped Fe 2 O 3 PVA thin films were also investigated. By using obtained results, the Linear Attenuation Coefficient (LAC), Half Value Layer (HVL), Tenth Value Layer (TVL), Mass Attenuation Coefficient (MAC). Also, theoretically Z eff value have been calculated. According to all obtained results, it has been seen that the different percentages (0, 5, 10, and 20%) Gd doped Fe 2 O 3 PVA thin films can be used as radiation shielding material. • Purpose of this study is to investigate effect of gadolinium (Gd) content of GdxFe2O3(100-x), (x=0, 5, 10, 15, 20) composite system on structural, thermal, optical and radiation shielding of poly (vinyl alchool) PVA based nanocomposite film. • The characterization of these powders have been obtained by using X-ray diffractometer (XRD), scanning electron microscopy (SEM), and Uv–vis spectrometer. • The irradiation properties of 6 MeV energized X-ray via cLINAC to obtain radiation shielding performance. • By using obtained results, the Linear Attenuation Coefficient (LAC), Half Value Layer (HVL), Tenth Value Layer(TVL), Mass Attenuation Coefficient(MAC). • Also, theoretically Zeff values have been calculated. [ABSTRACT FROM AUTHOR]

Ionizing radiation exposure from medical applications is increasing annually worldwide. It was estimated that 325 million dental procedures were performed in the United States. Radiation exposure from dental radiography consists of intraoral, panoramic, and 3D imaging cone-beam computed tomography (CBCT) imaging. Recent studies reported an association between dental imaging procedures and increased cancer probability of brain and thyroid. Previous studies showed that some dental imaging practices exposed patients and staff to unnecessary radiation doses due to incorrect image acquisition and insufficient radiation protection measures. This study aims to (i) measure the occupational and patients doses during dental procedures and (ii) assess the current imaging techniques and radiation protection practices. Two hundred fourteen patients were evaluated for periapical, bitewing, cephalometric, occlusal, and panoramic procedures. Organ equivalent doses were quantified for the breast, eye lens, and thyroid gland during CBCT procedure. Occupational and ambient dose assessment were assessed using calibrated thermoluminescent dosimeters (TLD-100(LiF: Mg. Ti). Ambient doses were measure at different locations at the department using TLDs. Patients' radiation doses were quantified using kerma area product (PKA (mGy.cm) and the entrance surface air kerma (ESAK (mGy). Fixed tube voltage (65 kVp) and tube current-time product (7 mAs) were used. The overall mean, sd, and range of patients dose values during intraoral (mGy), panoramic and CBCT examinations were 4.6 ± 0.7 (1.4–7.1), 135 ± 45 (75.2–168.5), and 215 ± 165 (186–2115), respectively. The mean and range of the annual occupational doses (mSv) were 1.4 (0.6–3.7), which below the annual dose limits for radiation workers (20 mSv/y). The study showed that inadequate radiation protection for patients existed in terms of the use of the thyroid shield, the technologist's presence inside the room during radiation exposure. Patients' radiation doses were comparable with the international diagnostic reference level (DRL). Staff education and training in radiation protection aspects are highly recommended. • Radiation doses from intraoral, panoramic, and 3D imaging cone-beam computed tomography imaging were quantified. • Ambient and occupational doses were measure at different locations at the department using TLD. • The study showed that inadequate radiation protection for patients existed in terms of the use of the thyroid shield. • CBCT dose values, the patient radiation dose is around two times higher than for a panoramic imaging procedure. • The mean and range of the annual occupational doses (mSv) were below the annual dose limits for radiation workers. [ABSTRACT FROM AUTHOR]

Abstract: Tincalconite, ulexite and colemanite are boron ores. Since these ores include boron, hydrogen and many other elements, these boron ores may be used as shielding materials instead of Pb metal and paraffin wax. In this study, measurements have been made to determine radiation transmission of tincalconite, ulexite and colemanite by transmission method for 6 and 18MV using linear accelerator (LINAC) with ionization chamber. The experimental results were compared to the results of WinXCom X-ray computer program’s. In pair production region, ores such as colemanite, ulexite and tincalconite can be preferred as shielding materials to prevent photon and neutron particle radiations. [Copyright &y& Elsevier]

Abstract: This paper presents results of equivalent dose organ determining to both primary beam and scattered radiation in a computed tomography (CT) procedures. All measurements were carried out to abdomen and critical organ (gonads) area. Selected dosimeters showed a standard deviation of 4.1% below to the reference values established by international guide lines. The equivalent dose in gonads was 14.27mGy. [Copyright &y& Elsevier]

Abstract: To provide medical surveillance of workers exposed to risk of internal contamination, IRSN (French Institute for Radiological Protection and Nuclear Safety) has developed two mobile laboratories for on-site monitoring. The laboratories are unique in Europe. They meet the new radiation protection requirements for nuclear medicine departments and radiological emergency response. Details of the design, calibration procedures and performance characteristics of these systems in measurements of various types of organs (thyroid, lung and whole body) are described. The sensitivity of the measurements is very close to that achieved in a heavily shielded stationary laboratory. [Copyright &y& Elsevier]

Abstract: A reliable determination of Technologically Enhanced Naturally Occurring Radioactive Materials in phosphogypsum is necessary to comply with radiation protection and environmental regulations. In this respect, a new phosphogypsum reference material was produced and certified to assist in the validation of analytical methods and the quality assurance of produced analytical results. This paper presents the sample preparation methodology, material homogeneity assessment, characterization campaign results and assignment of property values, and associated uncertainties. The reference values and associated uncertainties for Pb-210, Ra-226, Th-230, U-234 and U-238 were established based on consensus values calculated from analytical results reported by three National Metrology Institutes and five expert laboratories. [Copyright &y& Elsevier]

Abstract: [131I]-metaiodobenzylguanidine (mIBG) is a known radiopharmaceutical used for the treatment of neuroendocrine tumors. The development of therapeutic [131I]-mIBG doses at production level is highly challenging due to rapid product degradation and high radiation exposures to the production plant personnel. In the present work, a working module for the production of 10 doses (100mCi each) in a single operation was developed following copper (I) assisted isotope exchange. The labeled product complies with the pharmaceutical specifications suitable for in-vivo patient use. [Copyright &y& Elsevier]