The purpose of the study is to investigate changes in radon concentration in modern multi-flat energy efficient buildings during the first years after construction. The results of sample radon surveys conducted in several large Russian cities and the data obtained by means of radon-monitors in multi-apartment, multi-story buildings were analyzed. According to the results of the study, an inverse correlation was found between radon concentration and the age of the building at the time of measurement. In the groups of buildings of the same construction period, the geometric mean radon concentration depends significantly on the age of the building. Repeated measurements of radon time series in three apartments showed a significant decrease in radon concentration under the same meteorological conditions after seven years. The effective leakage area, estimated using radon time series analysis, increases significantly during the first years after building's construction. It can be concluded, that a systematic, gradual decrease in radon concentration occurs during the first years after construction. On average, indoor radon concentration decreases approximately twofold during the first nine years of building operation. The reduction of the radon concentration should be considered in context of the building energy efficiency as it gives evidences that, over time, the condition of the structural elements in buildings deteriorates, increasing the permeability of the envelope. For radiation protection purposes, it is important that high radon concentrations exceeding the WHO reference level of 100 Bq/m3 were obtained in new energy-efficient buildings during the first few years after construction.

For a long time, radiation safety specialists have been aware of computer codes designed to calculate human dose exposure from beta or gamma radiation. However, these products have a major drawback: they do not consider the radiation fields generated by extended beta radiation sources due to bremsstrahlung. Existing solutions are either not widely available or use outdated nuclear-physical constants and methods for calculating dosimetric characteristics. A computer code was developed specifically for calculating shielding against bremsstrahlung. It allows to estimate effective, equivalent, exposure dose rate, ambient dose equivalent, and air kerma rate from bremsstrahlung. It utilizes the ANS/ANSI 6.4.3 materials data and the last dose conversion factors from ICRP 116.

The article presents estimates of the annual effective doses received by residents of multi-story residential buildings constructed after the year 2000 and classified as energy-efficient. The assessments cover nine major cities across Russia and focus on external and internal exposure. The calculations are based on field measurements, including non-destructive assessments of the average concentration of 226Ra, 232Th and 40K and effective dose rate, as well as concentrations of radon and the equilibrium equivalent thoron concentration within contemporary high-rise buildings. The average annual effective dose from external exposure for the surveyed cities was found to be 0.27 mSv, while the internal exposure attributable to radon and its progeny amounted to 0.90 mSv, and that from thoron progeny was estimated at 0.31 mSv.

The aim of the research is to develop a new method for estimation of the annual effective dose of thoron decay products for the population living in multi-story buildings. Preliminary measurements of the annual average thoron equilibrium equivalent concentration were performed in Ekaterinburg city considering the uncertainties of the developed method. The results were extrapolated to nine Russian cities using a non-destructive gamma-spectrometry technique, along with thoron equilibrium equivalent concentration estimations. Assessed annual average effective dose from thoron decay products was 0.32 mSv year–1 and ranged from 0.12 to 0.63 mSv year–1 depending on the city.

In this study, we explore the interactions of photons, protons, and alpha with matter in the context of two drugs, MDPLu and EDTMPLu, delving into the interesting field of radiation shielding. Our research is focused on the mass attenuation coefficient (GMAC), effective atomic number (Zeff), buildup factors (EBF and EABF), and dosimetry parameters with the goal of understanding how these factors affect how well these drugs work as radiation shields. Our results show that the effective atomic number was higher in substances with a higher percentage of Z element component, and that the GMAC values for MDPLu medications were the highest, while the GMAC values for EDTMPLu medications were the lowest. Additionally, MDPLu showed the lowest exposure buildup factor (EBF) and energy absorption buildup factor (EABF), further validating its efficacy as a shielding material. Moreover, MDPLu displayed a higher mass-stopping power (MSP) for proton interactions in the lower energy range, making it a prime candidate for specific radiation shielding applications. These insights contribute to our understanding of the radiation-shielding efficacy of bone cancer treatment medications.

PURPOSE:Assessment of absorbed doses on organs and tissues of miners during radon exposure in the Schneeberg mines in the sixteenth century and calculation of the probability of occurrence of radiation-induced lung cancer and lung fibrosis, considering the life expectancy characteristic and the absence of smoking. MATERIALS AND METHODS:The expected radon concentration at the Schneeberg mines has been estimated using published data. Modeling of the accumulation of radon in the working tunnels of mine workings was carried out using the RESRAD-Build 4.0, based on the radium concentration in soil and geometric parameters of the mining tunnel from the engravings in Agricola's book. The dynamics of radionuclides in the human body were performed using the WinAct software in accordance with data from ICRP Publications 130 and 137. The values of absorbed doses on the tissues of the respiratory tract were obtained using the IDAC 2.1 program. Several models based on the epidemiology of uranium miners have been used to calculate radiation risks from radon exposure. The probability of male survival at birth and the age-specific frequency of spontaneous lung cancer not associated with radiation for miners of the sixteenth century (nonsmoking men aged 20-40 years) were estimated to properly calculate the radiation risks. RESULTS:The expected radon concentration in the Schneeberg mines was assessed in the range of 75-100 kBq m-3. The average value of the equilibrium factor was estimated as 0.49 ± 0.03. The annual exposure of miners to radon decay products was assessed as 125-165 WLM year-1. The annual values of absorbed doses to different sections of the respiratory tract were calculated, the maximum absorbed doses of α-radiation are formed on the bronchial and bronchiolar regions of the lungs (2.23 Gy year-1). The deterministic effects as radiation fibrosis of the lungs with 10 years of experience in the mines of Schneeberg have a probability of occurrence from 60 to 100%. All the models used for radiation risk assessments showed that the lifetime risk of developing lung cancer for nonsmoking Schneeberg miners is many times lower than the risk of developing deterministic radiation effects. In contrast, for the smoking cohort of miners in the nineteenth century lung cancer become the dominant cause of death. CONCLUSIONS:The deterministic radiation effects of Schneeberg miners in sixteenth century, exposed to extremely high levels of radon, such as radiation pneumosclerosis or pulmonary fibrosis, are more likely than the development of radiation-induced lung cancer.

The activity concentrations of 226 Ra, 232 Th and 40 K were determined in 43 soil samples collected from 0 to 10 cm and 10 to 50 cm depths of Niška Banja region, Serbia, using a low background gamma spectroscopy with 3″ × 3″ NaI(Tl) scintillation detector. The mean absorbed dose rate of 0–10 cm and 10–50 cm depth soil were 66.1 and 60.4 nGy/h, respectively which was close to UNSCEAR worldwide value. The radium equivalent and annual effective dose effect were lower than the permissible level. Therefore, there is no significant radiological risk from the soil to residents living in this region.

It has recently come to light that radiotherapy with Ra-223 (alpha emitter) radionuclides can be an effective therapeutic option for a variety of different cancers. A comprehensive knowledge of biodistribution, accumulation, and clearance pathways, as well as radiopharmaceutical kinetics, is required for an accurate assessment of therapy efficacy and a reduction in the severity of any adverse effects that the treatment may cause. Using the absorbed dose, measured in mGy/MBq, researchers can explore each organ in adult male and female animals and the differential accumulation, removal rates, and clearance pathways of radiopharmaceuticals. Because of their high linear energy transfer (LET) and limited range in tissue, the use of alpha emitter radionuclides in radiotherapy has attracted substantial interest recently. These properties allow for targeted therapy with minimum damage to the healthy cells in the surrounding area. This article provides a complete literature analysis on applying alpha emitter radionuclides in radiotherapy. The review focuses on zutilizing two well-known software tools, WinAct 1.0 and IDAC-Dose 2.1, for dosimetry calculations and treatment planning. The findings highlight the significance of tailored dose planning, organ-specific adverse effects, optimum treatment regimens, the potential for combination therapies, and the advancement of radiopharmaceutical development.

The objective of the study was to estimate the effective dose rates of external population exposure based on measurements of average specific activity of natural radionuclides ( 226 Ra, 232 Th and 40 K) in building materials of existing buildings. The measurements were performed using new developed non-destructive technique in 100 apartments in 9 Russian cities. The effective dose rate is 34 nSv h –1 in average and varies from 10 to 102 nSv h –1 between cities. Specific factors of conversion from specific activity to effective dose rate were obtained taking into account the real room geometries.

This study highlights the use of 89Zr-oxalate in diagnostic applications with the help of WinAct and IDAC2.1 software. It presents the biodistribution of the drug in various organs and tissues, including bone, blood, muscle, liver, lung, spleen, kidneys, inflammations, and tumors, and analyzes the maximum amount of nuclear transformation per Bq intake for each organ. The retention time of the maximum nuclear transformation and the absorbed doses of the drug in various organs and tissues are also examined. Data from clinical and laboratory studies on radiopharmaceuticals are used to estimate the coefficients of transition. The accumulation and excretion of the radiopharmaceutical in the organs is assumed to follow an exponential law. The coefficients of transition from the organs to the blood and vice versa are estimated using a combination of statistical programs and digitized data from the literature. WinAct and IDAC 2.1 software are used to calculate the distribution of the radiopharmaceutical in the human body and to estimate the absorbed doses in organs and tissues. The results of this study can provide valuable information for the biokinetic modeling of wide-spectrum diagnostic radiopharmaceuticals. The results show that 89Zr-oxalate has a high affinity for bones and a relatively low impact on healthy organs, making it helpful in targeting bone metastases. This study provides valuable information for further research on the development of this drug for potential clinical applications.

A series of glasses based on the nominal composition of (Na2O)(5) + (Al2O3)(10) + (SiO2)((85+x)/2) + (CaO)((85+x)/2) + (WO3)(x) glasses system were produced utilizing the usual melt quenching process in this study. Experimental techniques and the FLUKA Monte Carlo algorithm were used to examine the properties of silicon-calcium glasses containing tungstate-III-oxide. For five glass structures identified according to (Na2O)(5)+(Al2O3)(10)+(SiO2)((85+x)/2)+(CaO)((85+x)/2)+(WO3)(x) (0 <= x <= 20 wt-%) glass composition, the impact of tungstate-III-oxide with ratios of (0 <= x <= 20 wtpercent) on radiation shielding characteristics of glasses was set. The densities of the produced glasses fluctuated between 2.847 g/cm(3) and 3.122 g/cm(3) when tungstate-III-oxide was substituted. The produced sample densities, which are important in assessing radiation shielding features, rose as the WO3 concentration increased, according to our first results. In addition, the structure of each sample was studied using FT-IR. FT-IR showed that when WO3 levels rose, the connection level increased, and the FT-IR spectra shifted to higher wavenumbers. The synthesis of WO3 in a glass matrix enhances the structural network by raising oxygen levels, which leads to the transition of SiO2 into - CaO. Elastic moduli and Ultrasonic velocities were found to rise as the ratio of WO3 in the generated samples increased. These two approaches were used to model linear and mass attenuation coefficients, photons-transmittance versus photon energy, radiation protection efficiency against photon energy, and absorber thickness (experimental and simulation). Based on the results, it can be stated that the w20 sample, which contains 20 wt%, will play the most effective function in radiation shielding. Increases in WO3 led in considerable increases in linear and mass attenuation coefficient values, which directly contribute to the development of the glass's radiation shielding characteristics.

Through experimental and modeling techniques, this research sought to investigate the reflections of partial replacement of CaO–SiO2 with Tungsten(III) oxide and its effect on structural, optical, and physical properties. The melt quenching technique was used to produce several glass samples with a nominal composition of 5Na2O–10Al2O3–(42.5 − x)SiO2–(42.5 − x)CaO–xWO3 system (where x = 0, 0.2, 0.4, and 0.6 wt.&). The amorphous structure of calcium-silicon glasses was determined experimentally using the XRD technique. UV and density studies were also performed to determine optical and material properties. To determine the effect of this replacement on nuclear radiation shielding improvement, the linear attenuation coefficient was computed across a broad energy range of 0.015–15 MeV using narrow beam geometry and the simulated gamma-ray transmission technique. The radiation parameters were simulated using sophisticated Monte Carlo simulations using the FLUKA general-purpose radiation transport algorithm and compared using the NIST XCOm theoretical computation. To maximize the substitution's synergistic impact, the present investigation's findings were correlated with each other for the purpose of determining availability for nuclear shielding purposes. It was discovered that when the WO3 content rises from 0 to 20% wt— percent, both direct and indirect bandgaps reduce, resulting in increased transparency. In addition, the Urbach energy (Eu) yields a rise in proportion to the amount of doping elements in the samples. On the other hand, gamma-ray attenuation measurements revealed that shielding parameters are directly proportional to the WO3-wt percent concentration; furthermore, the addition of WO3 substantially enhances the shielding capacity of the investigated glasses against gamma radiation. It is observed that when additional WO3 is added, both the real and imaginary components of the dielectric constant increase. The most effective shield was determined to be the NSCW20 sample. As a result, a WO3-calcium-silicon glass containing 20% WO3 was selected as the optimum radiation filter.

The measurement of radon decay products level in dwellings or working places separately is not preferable. The estimation of radon equivalent equilibriwn concentration (EECRn) is more simple and quick technique. In this work, the uncertainty of calibration system for EECRn measurements will present and the reduction will be suggesting. The calibration system for EECRn measurements was presented and described in previous work with gamma spectrometer as a reference measuring device. The influence of alpha particles absorption in filters materials and filter efficiency taken into account. The measurements of EECRn by gamma spectrometry and improved alpha radiometry are in good agreement and the systematic shift between average values is observed and resolved. The total standard uncertainty of EECRn measurements with gamma spectroscopy is 3.8%. About 71% of this total uncertainty value is related to the uncertainty of the count rate at full absorption peak with gamma spectroscopy. If the time between the end of sampling and gamma spectroscopy measurements reduced to 2000s not 4000 s, this value will reduce to 1.6% and the total standard uncertainty of EECRn will be 2.6%.

In multi-storey buildings, Ra-226 in building materials is the main source of radon entry. Under low air exchange rate (e.g. in energy efficient buildings), radon concentration may exceed reference levels used for radiation protection purposes. Prevention and remediation measures of protection against indoor radon in situation when building materials are the main source of radon entry requires further development. The aim of this work is to evaluate the effectiveness of various radon prevention and mitigation measures in multi-storey buildings under different climatic conditions. On the basis of a dynamic model of indoor radon entry and removing, the following measures were simulated: control over of Ra-226 content in building materials, increasing uncontrolled air infiltration, extending the duration of controlled ventilation, equipping the building with mechanical ventilation, increasing the air exchange rate in non-living premises of the building, and reducing air exchange between living rooms and non-livings premises. The costs of energy for heating associated with different radon prevention and remediation measures were estimated. The conclusion about the potential effectiveness of various measures was drawn by achievement of optimized level of radiation protection, degree of reduction in the radon exposure risk, and cost of heating.

Long-term radon time series were analyzed in 14 rooms and offices to search factors influencing indoor radon concentration in high-rise buildings. Radon entry to living rooms is determined both by diffusion from building materials and advection from adjoined low-ventilated spaces. Humans affect on seasonal and diurnal radon concentration changes in high-rise buildings by activating ventilation system. Indoor-outdoor temperature difference significantly influence on the air exchange rate under uncontrolled ventilation, which is especially important in new energy efficient buildings.

A correlation between the mass concentration of particulate matter (PM) and the occurrence of health-related problems or diseases has been confirmed by several studies. However, little is known about indoor PM concentrations, their associated risks or their impact on health. In this work, the PM1, PM2.5 and PM10 produced by different indoor aerosol sources (candles, cooking, electronic cigarettes, tobacco cigarettes, mosquito coils and incense) are studied. The purpose is to quantify the emission characteristics of different indoor particle sources. The mass concentration, the numerical concentration, and the size distribution of PM from various sources were determined in an examination room 65 m3 in volume. Sub-micrometer particles and approximations of PM1, PM2.5 and PM10 concentrations were measured simultaneously using a diffusion aerosol spectrometer (DAS). The ultrafine particle concentration for the studied indoor aerosol sources was approximately 7 × 104 particles/cm3 (incense, mosquito coils and electronic cigarettes), 1.2 × 105 particles/cm3 for candles and cooking and 2.7 × 105 particles/cm3 for tobacco cigarettes. The results indicate that electronic cigarettes can raise indoor PM2.5 levels more than 100 times. PM1 concentrations can be nearly 55 and 30 times higher than the background level during electronic cigarette usage and tobacco cigarette burning, respectively. It is necessary to study the evaluation of indoor PM, assess the toxic potential of internal molecules and develop and test strategies to ensure the improvement of indoor air quality.

OBJECTIVE:In this article, IDAC-Dose2.1 and OLINDA computer codes are compared as they are the most widely used software tools for internal dosimetry assessment at the present time. OLINDA/EXM personal computer code was created as a replacement for the widely used MIRDOSE3.1 code. IDAC-Dose2.1 was developed based on the ICRP specific absorbed fractions and computational framework of internal dose assessment given for reference adults in ICRP Publication 133. IDAC uses cumulated activities per administered activity in hours and calculates the absorbed dose and the effective dose. The program calculates the dose in the Eckerman stylized family phantoms. It is useful in standardizing and automating internal dose calculations, assessing doses in clinical trials with radiopharmaceuticals, making theoretic calculations for existing pharmaceuticals, teaching, and other purposes.METHODS:To produce such a comparison, the results of this work were compared with available published data in the literature on radiopharmaceuticals. Radiopharmaceuticals with 89Zr, 153Sm, 177Lu radionuclides are used as the basis for the comparison. 89Zr, 153Sm, 177Lu radionuclides are regarded as the future of radiopharmaceutical treatment. For 89Zr, two different labelled carriers, Zr-89_cMAb U36 and Zr-89 Panitumumab, were used on patients.RESULTS:The results show a clear difference in terms of absorbed dose of the Zr-89 radiopharmaceuticals for red bone marrow when calculated by IDAC-Dose2.1 (0.76 mGy/MBq), while the estimated absorbed dose in literature results is 0.07 mGy/MBq and 0.14 mGy/MBq when the calculation is done by OLINDA program. In the case of 177Lu-EDTMP, the absorbed dose in red bone marrow is in reasonable agreement (0.63 mGy/MBq and 0.8 mGy/MBq for IDAC-Dose2.1 and OLINDA, respectively). A significant difference was found for the absorbed dose in the bone surface, which was almost twice as high for OLINDA (2.1 mGy/MBq for IDAC-Dose2.1 and 5.4 mGy/MBq for OLINDA). In some direct cases, the calculated absorbed dose in the urinary bladder wall with OLINDA is ten times higher compared to WinAct (which was utilized to calculate the total activity in the organs and tissues) and IDAC 2.1. These results are considered key to greater accuracy in internal dose calculation.

This study aimed to observe the reflections of partial substitution of boron trioxide with samarium (III) oxide and its impact on structural, optical and physical behaviors through experimental and simulation methods. Accordingly, a thallium-borate glass group with a nominal composition of 30Tl2O3+10Li2O+(60-x)B2O3+xSm2O3 system, where (x = 0, 0.2, 0.4 and 0.6 wt%), were prepared with the melt quenching method. The amorphous structure of thallium-borate glasses obtained using experimental XRD method. Further, UV and density analyses implemented to provide optical and material characteristics. A narrow beam geometry and experimental gamma-ray transmission method was used to measure the linear attenuation coefficient (LAC) at 0.081, 0.356, 0.662, 1.173 and 1.33 MeV to achieve the impact of this substitution on nuclear radiation shielding improvement. The radiation parameters were also measured using advanced Monte Carlo simulations by FLUKA general-purpose radiation transport code. It was found that both direct and indirect band gaps increase the value of eV as the Sm2O3 content increases from 0 to 0.6 wt—%, leading to more transparency. The most effective shielding properties was found for the BTLSm4 sample with x = 0.6 wt—%. Therefore, a Sm-borate glass containing 0.6 Sm2O3 was chosen as the best filter against radiation.

To control the specific activity of 226 Ra in building materials of operated buildings, a non-destructive in situ method consisted in measurements of gamma spectra with a fixed geometry of the detector position in a room is suggested. The calculation of the average specific activity of natural radionuclides in building materials is carried out by comparing the calculated flux density of unscattered gamma quanta normalized to the specific activity of 100 Bq/kg, and the experimentally measured count rates in the photopeak. For the measured values of the average specific activity of 226 Ra in building materials, the expected values of radon concentration in rooms were calculated. The developed method has been verified for rooms in modern multi-storey energy-efficient buildings. Concentrations of 232 Th and 40 K in building materials which contribute to the indoor external radiation exposure can be measured by developed method also.