Health physics最新文献

In recent years, the use of radiation diagnostics, treatment services, and many areas of health services has become widespread with technological developments. The widespread use of radioactive substances and radiation-producing devices in health services has increased the need for effective radiation protection programs, and the need for training in terms of the applicability of these programs is also increasing. This randomized controlled trial was conducted to evaluate the effect of on-the-job toolbox training defined as a short, informal safety meeting led by a supervisor that focuses on specific workplace hazards or safe work practices on the knowledge, practices, and safety-related behaviors of healthcare professionals working in radiation environments. Sixty-three participants from the radiology units of a university hospital in Hatay, Türkiye, were randomly assigned to an intervention group (n=33) or a control group (n=30). The intervention group received brief, face-to-face training sessions using visual materials in their work environments. Data was collected before and after the training and at two follow-up periods. After the training, radiation protection knowledge increased significantly in the intervention group (p<0.001), while no change was observed in the control group. Observational assessments revealed that there was an improvement in practice scores in the intervention group and that the gains were largely maintained over time. Additionally, adverse event reporting, an important quality indicator, increased significantly in the intervention group, indicating increased safety awareness. These results confirm that toolbox training is an effective method for improving both knowledge and safe practices among healthcare professionals. Its brief, practical, and workplace-based format contributes to increased engagement and retention of learning. The findings support the integration of toolbox training into in-service training programs as a complement or alternative to traditional methods. Future studies should examine its long-term effectiveness and applicability in a variety of healthcare settings.

Evacuation and sheltering-in-place are key protective actions during the early phase of a radiological emergency. Protective action strategies and decisions require balancing risks to ensure they provide more benefit than harm. While evacuations are generally safe, there are deterministic, long-term health consequences from prolonged displacement that outweigh the stochastic radiological health risk. In many cases, sheltering-in-place provides a viable alternative to evacuation, yet guidance and tools are lacking to aid decisionmakers on the best choice of action during an emergency caused by a nuclear power plant. Therefore, a method was developed to evaluate the effectiveness of sheltering-in-place and to explore ways to make better use of shelters during a radiological emergency. A model to estimate the radiological protection afforded by a typical residential shelter was developed and incorporated into a tool to examine the effectiveness of shelters during a hypothetical severe accident for typical BWR and PWR reactor designs and potential small modular reactor (SMR) design. The resultant analysis tool provides valuable insights into the parameters important to effective sheltering. Results demonstrate that typical residential homes can provide substantial dose reduction for an extended period of time following a significant radiological release. A sensitivity study was performed to identify important parameters and to inform how to implement sheltering-in-place for radiological emergencies involving different types of reactors and accidents with different radiological source terms. The model and results compared well to similar models and studies that also suggest shelters are more protective than previously understood. Additionally, contrary to the conventional shelter-in-place strategies, which recommend securing building ventilation, the results of this study suggest that sheltering strategies may benefit from a judicious operation of heating and ventilation systems for use in purifying indoor air to reduce inhalation dose.

The current investigation involved collecting and analyzing groundwater samples from the Chengalpattu district in the state of Tamil Nadu. The content of uranium was determined using an LED fluorimeter. The uranium concentration exhibited a range of 0.02 μg L -1 to 2.814 μg L -1 , with a mean value of 0.347 μg L -1 and a standard deviation 3.05 μg L -1 . The uranium concentration levels calculated in this study are significantly below the recommended limits set by various agencies, including WHO (2011), US EPA (1991), ICRP (1995), UNSCEAR (1982), and AERB (2004), and these limits are used for critical in protecting public health. The annual ingestion dosage from uranium consumption via drinking water is estimated for all age groups and compared to the prescribed limits. The annual uranium intake in drinking water for various age groups ranges from 0.0224 μSv y -1 to 7.0674 μSv y -1 . The analysis shows that the doses are far below the recommended safety threshold. The average yearly intake dosage is well below the required threshold, indicating that the drinking water sources in the study region do not pose any harmful health risks related to uranium.

Internal dosimetry is a part of radiation safety for patients and radiation workers in nuclear medicine procedures. This study retrospectively determined the effective dose for oncology patients undergoing PET/CT scans by using widely used computer codes, and the results were compared with each other and literature. The study focused on the radiopharmaceuticals 18 F-FDG (n = 220) and 68 Ga-PSMA (n = 85), administered to 305 patients for cancer imaging at Yeditepe University Kosuyolu Hospital's nuclear medicine department. PET dose was calculated using OLINDA/EXM, IDAC-Dose 1.0 and IDAC-Dose 2.1 programs while ImPACT software was used to determine the effective dose from the CT scan. All effective doses were derived in accordance with ICRP 60 and ICRP 103 tissue weighting factors. PET effective doses from highest to lowest were calculated with OLINDA/EXM, IDAC-Dose 1.0, IDAC-Dose 2.1 (ICRP 60) and IDAC-Dose 2.1 (ICRP 103) as 9.96, 9.07, 7.01, 6.28 mSv respectively for 18 F-FDG. The highest PET effective dose was also calculated with OLINDA/EXM software as 3.65 mSv for 68 Ga-PSMA. For the total effective dose in PET/CT scans, CT contributed about 92% for 68 Ga-PSMA protocol and 75% for 18 F-FDG protocol.

This study investigates the impact of heating rates, ranging from 1 °C s- 1 to 20 °C s- 1 , on the precision of integrated peak counts determined using various thermoluminescent dosimeter materials. Lower heating rates influence precision due to prolonged integration of signal noise, while higher heating rates affect precision by pronounced thermal quenching effects. Using time-temperature profiles constructed with a linear heating ramp and a constant hold at maximum temperature, a range of heating rates was evaluated to identify an optimal condition that minimizes variance in integrated peak counts resulting from these effects. In addition, kinetic parameters of glow peaks were determined through peak deconvolution of each glow curve obtained and analyzed as a function of heating rate, with observed trends fit to appropriate models. These results were then compared to trapping parameters - namely the activation energy and frequency factor - independently extracted using the variable heating rate method to assess consistency across techniques. The results indicate that peak temperatures and intensities exhibit strong exponential dependence on heating rate, while activation energies and frequency factors show weak linear correlations. Trapping parameters obtained using the variable heating rate method fell within the range of values derived from peak deconvolution, supporting consistency between the two approaches. An optimal heating rate of 4 °C s- 1 was identified for minimizing variance in integrated peak counts across all dosimeter types tested. Both noise and thermal effects were shown to influence measurement variance, with thermal quenching effects having a more pronounced impact at higher heating rates. Additional factors affecting precision included dosimeter material, glow peak temperature, and overall glow curve complexity. These findings enhance the understanding of thermoluminescent dosimeter behavior and highlight the importance of optimizing the heating rate for improved measurement reliability.

Following the reduction in the occupational eye lens dose limit by the International Commission on Radiological Protection (ICRP), a physical phantom was developed to enable lens dose monitoring in complex radiation environments. The design is based on a validated eye model used to derive lens dose conversion coefficients. After evaluating various materials using Monte Carlo simulations, polymethyl methacrylate (PMMA) was selected for its ease of manufacturing and decontamination. The phantom's dose conversion coefficients were calculated for a range of photon energies and irradiation angles. Lens doses were measured using thermoluminescent dosimeter (TLD) chips embedded in the phantom and exposed to a 137 Cs reference field. Measurements were compared with simulation results, demonstrating good agreement. This newly developed PMMA phantom offers a practical solution for evaluating lens dose in non-uniform radiation fields such as those found in nuclear power plants.

We have developed a method for calculating the probability of doses during clearance of land that makes no a priori assumptions about the statistical distribution of radionuclide measurements. Our method uses the actual distribution of survey and sampling measurements rather than assuming a specific distribution. It applies Monte Carlo simulations to the Sum of Fractions calculation and accounts for all known uncertainties. In addition, it defines the probability and uncertainty of complying with the clearance criterion. In contrast, nearly all demonstrations of compliance with clearance criteria currently are based on nonparametric statistics. As such, the analysis is for the median dose to an individual in the critical group. They are based on an assumed statistical distribution and do not account for all known uncertainties. In contrast, regulatory authorities base compliance on either the dose to representative person or the dose to the average member of the critical group - not the median member of the critical group. Accounting for uncertainties is required as noted in the assertion of at least eight international authorities, including the International Organization for Standardization and the International Union of Pure and Applied Chemistry. They state that a measurement without its stated uncertainties is incomplete, not technically sound, and may not be considered defensible. Our Monte Carlo Sum of Fractions method provides the highly accurate quantified probability and uncertainty of compliance with the clearance criterion because we simulate all the data with randomized uncertainties. Therefore, it improves verification of clearance. In this work we demonstrate the application of Monte Carlo Sum of Fractions method to measurement data from a real site as a "Proof of Concept" for clearance of land. The evaluation of the dose to either of the representative person or the mean dose, and quantified uncertainty due to known parameters are technically sound basis for demonstration of compliance with clearance criteria.

Accidental trace 131I skin contamination resulted in an intake to a veterinary nuclear medicine technician at Colorado State University (CSU). The resulting dose to the technician was determined through an in-vivo measurement of radioactivity in the thyroid. Inhalation and ingestion of the isotope were ruled out due to the chemical nature of the 131I, as determined by a previous study. The CSU Radiation Control Office performed measurements of the technician's thyroid daily and then weekly to quantify the uptake [2.2 kBq (60.1 nCi)] as well as the effective half-life of 131I (7.61 days) for the technician. Establishing a technician-specific effective half-life by graphing in-vivo measurements also showed near agreement with established effective half-life determinations specified in ICRP 30 for inhalation. The university assessed the committed dose equivalent (CDE) to the thyroid using three different methods and assigned a the highest CDE calculated (4.1 mSv), and an SDE of 0.22 mSv from external contamination. The average of the CDE calculated via each method was 3.68 ± 0.38 mSv. A comparison of the 3 thyroid dose calculation methods resulted in a 10% coefficient of variation. Close agreement between the various calculations demonstrates that any of these methods would be sufficient in determining a committed effective dose to the thyroid, while providing a level of confidence to the technician that the determination is accurate and appropriate. A key aspect of this report is how even trace amounts of radioiodine on intact skin can result in measurable doses.

Objective: To analyze stable chromosomal aberrations and reconstruct the biological dose in 6 victims of the Henan "4.26" 60Co radiation accident, 9 and 20 years after irradiation.

Methods: G-banding chromosomal specimens were prepared using conventional culture and trypsin digestion methods. Stable chromosomal aberrations in six victims were analyzed with an automated chromosome scanning system and karyotype analysis software to estimate the biological dose based on chromosomal translocation indicators.

Results: The number of stable chromosomal aberrations per cell at the two follow-ups demonstrated a significant dose-effect relationship with the radiation dose. A noticeable increase in translocations per cell was observed 20 years after the accident compared with 9 years. At 9 years post-irradiation, the mean absorbed doses for four cases, estimated from translocations per cell, were similar to those estimated from dicentrics plus rings at 5 days post-irradiation, while a slight decrease was observed in one case. At 20 years post-irradiation, the mean absorbed dose based on translocations per cell for five cases was essentially the same as the dose estimated from dicentrics plus rings at 5 days post-irradiation.

Conclusion: The G-banding approach, characterized by its simplicity and low cost, is feasible for analyzing chromosomal translocations and retrospectively reconstructing the biological dose of early radiation accidents.

Radiation-based medical techniques and devices provide significant benefits to patients through the diagnosis, treatment, and management of illness and disease. Documenting trends and frequency of use offer important insights into radiation protection and help address gaps in the documentation of medical exposures. Here, we present the retrospective Canadian data collected for the recent United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) global survey on medical exposure. The global survey included three modality categories: diagnostic and interventional radiology, nuclear medicine, and radiotherapy and reports the total number of devices, physicians, examinations, and procedures. Due to the inability to collect high-quality dose data from Canadian sources, the average doses for specific examinations and treatments were estimated using internationally pooled data. The total annual per capita dose from medical exposures was determined to be 1.56 mSv, excluding radiotherapy, resulting in approximately 47% of all radiation doses received by Canadians, compared to natural, industrial, and consumer product sources. This assessment of Canadian medical radiation exposures contributes to global improvement of patient protection, helps establish trends, and identifies where Canadian data collection is lacking, particularly dose data.