Automation limits operator contact during routine 18F-fluorodeoxyglucose (FDG) production, but residual manual tasks and nonroutine interventions may still determine occupational exposure. This study quantified workflow-specific effective dose during in-house FDG production at a hospital-based cyclotron facility. Production records from 39 batches performed from 4 July to 23 August 2024 were reviewed. Two radiological technologists were monitored during seven routine production sessions, and five pharmacists were monitored during 10 quality-control (QC) observations. Semiconductor electronic pocket dosimeters were worn at the head, chest, and abdomen; effective dose was estimated from the three 1-cm dose equivalent readings using a nonuniform exposure formula. For technologists, average effective dose was 0.4 ± 0.1 µSv per session for cyclotron operation and synthesis, 0.2 ± 0.1 µSv for transport, 1.1 ± 0.3 µSv for cleaning, and 1.6 ± 0.3 µSv for the total routine workflow. The average pharmacist QC dose was 2.5 ± 1.2 µSv per observation. Two of 39 batches (5.1%) required a nonroutine intervention. Event-related effective doses were 32.2 µSv for manual recovery after QC vial-transfer failure and 30.7 and 31.6 µSv for two operators recovering residual FDG after incomplete transfer. These values were approximately 19-20 times the average total routine technologist dose. Routine exposure was low, whereas infrequent disruptions produced disproportionate exposure. Radiation protection should therefore combine routine monitoring with prevention, rehearsal, shielding, and time-minimized procedures for manual QC and nonroutine recovery.

The Council on Ionizing Radiation Measurements and Standards (CIRMS) is an independent, non-profit council composed of experts from academia, government, and industry across all areas of ionizing radiation science and technology, providing leadership, focus, and coordinated action on emerging radiation metrology needs. This mission is achieved through the organization of an annual scientific conference and the periodic publication of a Needs Report which identifies priority areas in ionizing radiation measurements and standards. This article summarizes the outcomes from the CIRMS 2026 meeting which was held 13-15 April 2026 at The Universities at Shady Grove in Rockville, Maryland, United States. Participants represented a broad cross-section of the radiation science community, including radiation metrologists, academic researchers, government scientists, and industry stakeholders. The meeting featured plenary presentations, topical breakouts, and poster sessions. The breakout sessions were organized by the three subcommittees of the CIRMS Science and Technology Committee: Medical Applications, Radiation Protection and Homeland Security, and Radiation Processing and Material Effects. Topics covered during the meeting included cobalt-60 (60Co) and cesium-137 (137Cs) production, medical device sterilization, FLASH radiotherapy dosimetry, neutron dosimetry, radiation standards for next-generation fusion power systems, Monte Carlo modeling, and other topics relevant to radiation measurements and standards. The CIRMS 2026 meeting highlighted challenges that must be addressed to advance ionizing radiation science and technology and provide important input for the next Needs Report.

The general guidance for a radiological impact using pre- and post-assessments for legacy, new, and planned uranium mining facilities are presented. Specific items to consider during the process and the rationale such as mitigation, order of model, loading functions and approaches, hydrographs, pollutographs, water quality modeling, and health physics are provided. A case study using a conventional Coles Hill proposed uranium mine in Virginia demonstrates typical pollutograph construction processes for a worst-case scenario 100-y intense rainfall pre-assessment. For the case study, an intense 3-h duration, 100-y synthetic rain event overlapping a sudden rupture of a full 5.00 × 104 m3 leachate (= 2.78 × 105 Bq m-3) storage tank over a 3-h duration was considered. Having similar characteristics, a surrogate Rio Road Watershed hydrograph was used to emulate the local watershed at White Horn Creek, Coles Hill. The uranium mine waste (leachate, slime, sand) was transported over the local land area and river embankment into the nearby Banister River as it mixed with a spatial temporal runoff rate (q-Series). A Continuous Spill/Step Loading Runkel water quality model determined the concentrations in 15-min increments the entire length of the Banister River downstream to the Kerr Dam. There was insignificant evidence that there would be an exceedance of the MCL for 226Ra/228Ra (= 1.85 × 10-2 Bq m-3), total alpha (= 5.55 × 10-2 Bq m-3), or uranium (= 3.00 × 10-6 Kg m-3) at the Town of Halifax's drinking water intakes (approximately 50 km). Therefore, it is highly unlikely that there would be an exceedance of 226Ra/228Ra, total alpha, or uranium near the drinking water intakes of the Town of Halifax, Virginia, or further downstream at Clarkesville's and Kerr Dam's drinking water intakes. Furthermore, there would be a need for post-disaster data collection and potential remedial action of radioactive contamination in water, soil, sediment, building structure, and the ecosystem.

The present study investigates uranium concentrations in 345 drinking water samples collected from the Dausa and Sawai Madhopur districts of Rajasthan, India, and assesses the associated radiological and chemical health risks. The measurement of uranium concentrations was performed using the light-emitting diode fluorimetric method (Model LF-2a). The uranium concentrations vary between 0.52 and 222 µg L-1, with an arithmetic mean of 14.8 µg L-1, and a median 7.7 µg L-1. This investigation found that 12.2% of the samples exceeded the World Health Organization guideline value of 30 µg L-1. The assessment of radiation dose considered the differing water intake patterns among various age groups, indicating that infants are the most vulnerable to uranium-induced health impacts. The average excess cancer risk obtained in the present study is lower than the value prescribed by India's Atomic Energy Regulatory Board (1.67 × 10-4). The results indicate that the mean daily intake of uranium varied between 0.01 and 6.33 µg kg-1 d-1, and hazard quotient analysis indicated potential chemical toxicity risks in 9.6% of the samples. Uranium was found to be positively related to total dissolved solids.

One of the main concerns of radiography procedures is to produce diagnostic image quality with dose reduction, specifically for pediatric patients. In the present study, Monte Carlo simulations using GATE (Geant Application for Tomography Emission) code, coupled with a 7-wk-old voxel female baby model from the GSF (German Research Center for Environment Health) were performed to investigate how dose in digital chest, abdomen, and both chest and abdomen (babygram) x-ray examinations vary with tube voltage (40-60 keV) at a fixed value of current-time product (4 mAs) using different collimation and a digital radiography system. Organ dose (mGy) and effective doses (mSv) of three routine digital radiography examinations, namely anterior-posterior (AP) chest, anterior-posterior (AP) abdomen, and anterior-posterior (AP) chest-abdomen, were estimated using the Monte Carlo simulation method, GATE code. Organ doses ranged from 0.01 to a maximum of 2.5 mGy, while effective doses ranged from 0.01 to 0.3 mSv. The results of the present paper show that the organ doses and effective doses for chest, abdomen, and babygram x ray with digital radiography highlight no change, while increasing voltage and values obtained are still very low compared to similar references.

To estimate the annual population dose from medical examinations, the frequency of the examinations and the effective dose (E) for each examination needs to be estimated. In the last Swiss survey covering 2023, the dose per capita due to medical irradiation was estimated to be 1.7 mSv. In the framework of the 2023 survey, the E for x-rays' modalities were calculated using different methods (bibliographic review, data from a dose management system, data from older surveys, etc.). This work aims to present the methodology followed for the estimation of E for various modalities. E was estimated for approximately 60 radiological examinations and groups of examinations. The advantages and limitations for each method are discussed.

 This survey aimed to estimate the radiation exposure of the Swiss population from medical diagnostic procedures performed in 2023. The survey was based on billing data and administratively collected data for radiology and nuclear medicine and on data collection for the frequency of dental examinations. A total of 12.9 million radiological examinations were performed, corresponding to 1,443 examinations per 1,000 inhabitants. The collective effective dose was 1.69 mSv per capita, presenting a 13% increase compared with the dose estimated in 2018. Despite accounting for only 14.4% of the total number of examinations, computed tomography accounted for 75.7% of the total dose. In 2023, 207 CT scans per 1,000 inhabitants were recorded, with an average dose of 1.28 mSv per capita. The highest contribution to the examination frequency (46.5%) was recorded in dentistry, but this accounted for only 0.6% of the total dose (0.01 mSv per capita). The survey resulted in an update of the national data on population dose exposure from medical diagnostic procedures.

Radium-bearing naturally occurring radioactive material is a recurring radiation safety concern in oil and gas production. The radionuclides 226Ra and 228Ra may be mobilized in formation water, transported with produced water, and concentrated in scale, sludge, filters, tanks, separators, pigging residues, soils, and sediments. This paper proposes a semi-quantitative occurrence susceptibility index for radium-bearing scale and sludge to inform radiation safety survey planning. The framework combines five weighted components: radium mobilization potential, scale precipitation potential, produced-water burden, deposition and accumulation potential, and operational trigger potential. Empirical radiological evidence is treated separately through a noncompensatory trigger so that an elevated field indication or confirmed radium-bearing material cannot be diluted by low scores in other components. A data quality index reduces false confidence when information is incomplete. Four worked operational scenarios demonstrate how the core score, empirical trigger, and data quality alter survey sequence, portable instrument use, targeted sampling, high-purity germanium gamma-spectrometric confirmation, and work controls. The framework is a rule-based decision-support structure, not a validated predictive model, probability estimate, or substitute for direct measurement, regulatory requirements, or site-specific dose assessment. No field-validated improvement in efficiency or cost savings is claimed. Its present purpose is to make provisional survey planning transparent and testable until field calibration and validation are completed.

Radioactive air emissions are regulated by both the US Environmental Protection Agency (US EPA) and the State of Washington. Both agencies have adopted an annual dose limit of 0.1 mSv y-1 (10 mrem y-1) effective dose equivalent to a maximally exposed individual (MEI) member of the public. Compliance to the US EPA standard is determined by calculating the highest dose from radioactive air emissions due to point source emissions to a member of the public at a residence, school, business or office. Under the Washington State requirements radionuclide emissions and doses to members of the public include not only point sources but also radon and diffuse/fugitive doses. This retrospective review summarizes the annual Hanford Site radionuclide air emissions reports for the past 35 years beginning in calendar year 1990 through 2024, and it compares the results of the federal versus the state reported doses to the MEI. Additional considerations include the dose model version, location of the reported MEI, maximum emission unit contributor to dose, and the percent contribution of the predominate emitted isotopes. The comparison of the results from the two reporting methods shows low doses, in comparison to the limits, to members of the public and that the more expansive state doses are typically 2.4 times more than that of the federal requirement. This variation appears to be independent of the CAP88 dose model used. Finally, the annual radioactive air emissions dose to the MEI from the Hanford Site over the 35-year period to a member of the public is on average about 1% of the 0.1 mSv y-1 state limit.

Deployable radiological laboratories have evolved substantially over the past 20 years in response to the need for rapid, medically interpretable information during radiological emergencies occurring far from fixed analytical facilities. Early French mobile laboratories were limited to qualitative NaI(Tl) surveys and relied heavily on centralized laboratories for confirmatory analysis, resulting in delayed dose assessment and limited forensic capability. Building on operational experience and NATO doctrine (STANAG 4590), a four-module deployable system - the Radiological Emergency Laboratory (Laboratoire d'Urgence Radiologique, LUR) - was developed to integrate in vivo spectrometry, radiotoxicology, physical dosimetry, and cytogenetic triage within a unified operational architecture. The system can be mounted on armored Griffon vehicles or 4 × 4 all-terrain platforms, enabling deployment in operational environments where civilian mobile laboratories cannot operate. Modules 1 and 2 provide rapid screening and confirmatory in vivo measurements, while Modules 3 and 4 support comprehensive radiotoxicology and dosimetry. Analytical performance (ICP-MS, α-spectrometry, HPGe spectrometry) approaches that of fixed laboratories, enabling isotopic discrimination and early estimation of committed effective dose. By resolving most cases directly in the field, the LUR reduces the number of samples requiring referral to central radiochemical laboratories, thereby preventing saturation of national analytical capacity during large-scale radiological incidents. The LUR therefore provides a medically oriented, operationally robust, and economically sustainable capability for modern radiological emergency response, ranging from localized contamination events to large-scale population triage.

In this study, naturally occurring radionuclide concentrations and their respective radiological hazard indices in water samples collected from the eastern area of Lake Hawassa were measured. Analysis was conducted using HPGe gamma-ray spectrometry for the radionuclides 238U (214Pb and 214Bi), 232Th (228Ac), and 40K. Activity concentrations were found to range between below the detection limit (BDL) and 4.2 ± 0.8 Bq L[-1] for 238U, BDL and 9.1 ± 1.1 Bq L[-1] for 232Th, and BDL and 92.1 ± 4.5 Bq L[-1] for 40K with averages of 2.4 ± 0.6, 6.1 ± 0.6, and 55.1 ± 2.8 Bq L-1, respectively. Radiological hazard indices were determined and found to be the external hazard index (= 0.04), absorbed dose rate (= 7.14 nGy h[-1]), and annual outdoor effective dose equivalent (= 0.009 mSv y[-1]). These values were compared with international guidelines and other countries' data on natural radioactivity. From the comparison, it is evident that the detected radionuclide activity concentrations and their respective radiological hazard indices comply with internationally recommended limits, implying that the lake water poses no significant radiological risks to the local population. In addition, the gathered information is used as initial reference data but not as the final baseline value for Lake Hawassa regarding radiological hazards. These data will play an important role in establishing future environmental policies and programs for the study area.

In the event of worker contamination, and because dissolution properties of aerosols highly influence committed effective dose assessment, information regarding the bioavailability of inhaled contaminants is of great interest to the person in charge of the exposed workers. In this study, we developed and validated a new procedure to evaluate the early bioavailability of actinide compounds obtained at the workplace. Using the proposed method, actinides recovered from swipes taken in different glove boxes were analyzed using an in vitro multicompartmental and dynamic assay to assess the proportion of bioavailable elements. The transferred activity from the static phase to the dynamic phase of the assay device represents the bioavailability of the compound studied. Our data demonstrate the variability in the dissolution properties of the collected compounds among the different glove boxes with no clear relationship with their isotopic and elemental composition. We confirmed the influence of the presence of bioligands or synthetic chelators, as well as lung fluid surrogates, on the bioavailability of the compounds. Using a new method, intentionally designed to be quick and simple, we confirmed that each glove box contains compounds with specific dissolution properties. Therefore, our process deserves to be part of a toolbox for risk and dose estimation following accidental actinide intake at the workplace.

A pocket-sized dosimeter/spectrometer with 1-cm3 cerium-doped gadolinium aluminum gallium garnet, GAGG(Ce), scintillation crystal was evaluated for its application in radioactivity measurements. The device was used (1) for the measurement of in situ ambient gamma dose rates, (2) as a car-borne dosimeter to measure dose rates along the roads, and (3) for recording the energy spectra from different environmental samples by placing it inside the lead shield. The dosimeter showed relative uncertainties better than 10% for dose-rate measurements of about 30 nSv h-1 and higher when measurements were performed for 30 min or longer. As a spectrometer, it revealed the typical features of environmental gamma-ray spectra. Energy spectra from radon charcoal canisters exposed to different indoor radon concentrations demonstrated the capability of the device to study variations in radon progeny gamma-ray spectra.