Ambient dose equivalent rate and soil contamination density with 137Cs in kitchen gardens in settlements of the Bryansk region, Russia in 2020–2021

The article provides results of application of the field (in situ) gamma spectrometry method for carrying out mass monitoring measurements of ambient dose equivalent rate and soil contamination density with 137Cs in kitchen garden plots located in the zone of radioactive contamination after the Chernobyl accident. In 2020 and 2021, 115 private farmsteads in 46 settlements of the Bryansk region were surveyed. At the time of the survey, the officially established average density of soil contamination with 137Cs in the settlements ranged from 27 to 533 kBq/m2 . The field spectra were measured using a portable scintillation gamma-spectrometer-dosimeter. Results of the field measurements and subsequent calculations of soil contamination density with 137Cs in the kitchen gardens were in good agreement with official data on the average soil contamination density with 137Cs in the surveyed settlements. The mean value of the ratio of the experimental data to the official data was 1.04. Individual values of experimental data deviated from corresponding official values by no more than two times. The use of the gamma spectrometry method in situ made it possible: 1) to determine separately values of the ambient dose equivalent rate from 137Cs and from natural radionuclides in the soil, and 2) to estimate the effective external doses to a person who worked in the kitchen gardens. The measured values of ambient dose equivalent rate varied from 17 to 53 nSv/h (mean ± standard deviation = 35 ± 9 nSv/h) for natural radionuclides and from 8 to 432 nSv/h (mean ± standard deviation = 125 ± 91 nSv/h) for 137Cs. The ambient dose equivalent rate from 137Cs normalized to the soil contamination density with 137Cs in the same kitchen garden was in the range of 0.41–0.84 (nSv/h)/(kBq/m2 ) with a mean value of 0.55 (nSv/h)/(kBq/m2 ). If a person stayed in kitchen garden for 840 hours per year, the estimated effective external doses from natural radionuclides and 137Cs were respectively in the range of 0.008–0.025 mSv/year and 0.004–0.20 mSv/year.

Brief Report: Higher Fentanyl Exposures Require Higher Doses of Naloxone for Successful Reversals in a Quantitative Systems Pharmacology Model

Background: Previously, we reported on an opioid receptor quantitative systems pharmacology (QSP) model to evaluate naloxone dosing. Methods: In this study we extended our model to include higher systemic levels of fentanyl (up to 100 ng/ml) and the newly approved 8mg IN naloxone dose (equivalent to 4 mg)Results : As expected, at the lower peak fentanyl concentrations (25 ng/ml and 50 ng/ml), the simulations predicted that 2 mg, 4 mg, 5 mg, and 10 mg IM doses of naloxone displaced fentanyl and reached below the 50% receptor occupancy within 10 minutes. However, at the concentration of 75 ng/ml, the simulation predicted that the 2 mg dose of naloxone failed to reach below the 50% occupancy within 10 minutes. Interestingly, at the highest peak concentration of fentanyl studied (100 ng/ml), the model predicted that the 4 mg of naloxone IM (equivalent to 8 mg IN) failed to reach below the threshold of 50 % occupancy within 10 minutes or even within 15 minutes (Data not shown). In contrast, the model predicted successful reversals when 5 and 10 mg IM doses were utilized. Conclusion:These results support the notion that acutely administered higher doses of naloxone are needed for rapid and adequate clinical reversal, particularly when higher systemic exposure of the potent synthetic opioids occur.

ASSESSMENT OF RADIATION DOSE RATE LEVELS AND RADIATION RISK AT THE COBALT -60 UNIT, KOMFO ANOKYE RADIOTHERAPY CENTER, GHANA.

Purpose: A study was conducted to estimate the Annual Effective Dose Equivalent (AEDE) and Excess Lifetime Cancer Risk (ELCR) caused by the presence of an artificial cobalt-60 radioactive source producing ionizing radiation levels within the radiotherapy facility at Komfo Anokye Teaching Hospital (KATH) in Ghana. This study validated the safety of cobalt-60 radioactive sources, as well as the notion of calculating the Annual Effective Dose Equivalent (AEDE) and Excess Lifetime Cancer Risk (ELCR), which contributed to reducing occupational and public exposures inside the facility. Methodology: The investigation was carried out with the use of a portable OD-01 Ionization Chamber Survey Meter. The absorbed dose rate (ADR) in air was changed between 5 m and 40 m, with measurements taken inside and around the cobalt 60 bunker, as well as at sixteen other sites within the radiation facility. Findings: From 5 m to 40 m surrounding the Cobalt-60 source, the estimated Absorbed Dose Rate in air inside the cobalt-60 bunker ranged from 0.299 0.001 to 0.977 0.005 Sv/h, with an average of 0.498 0.005 Sv/h. The estimated Annual effective dose equivalent varied from 1.100 mSv/yr to 3.595 mSv/yr around the cobalt-60 source inside the Co-60 bunker. Radiation exposure levels ranged from 0.268 0.008 Sv/h to 0.678 0.005 Sv/h, with an average of 0.440 0.004 Sv/h observed around the fifteen sites chosen. Excess Lifetime Cancer has values ranging from 3.85 10-3 to 12.58 10-3 and 3.45 10-3 to 8.73 10-3. Risks were evaluated for the cobalt and the sixteen places inside the plant. The absorbed dose values at 5 m, 10 m, and 15 m inside the Co-60 bunker and the location Co-60 bunker as part of the facility exceeded the ICRP-recommended limit of 0.57. The AEDE and ELCR levels were within the ICRP's acceptable limits. The AEDE and ELCR statistics acquired indicate that the Cobalt-60 unit and its surroundings are radiation safe, although the likelihood of employees contracting cancer from the absorbed dose and background ionizing radiation is significant over a lifetime. Recommendation: However, it is recommended that absorbed dose level monitoring and evaluation of the Radiation Therapy Technologist (RTT) and other workers surrounding the unit be monitored on a regular basis. It is also recommended that Occupational Staff, such as RTTs, spend as little time as possible in the bunker

Correction Coefficients for Measuring the Ambient Dose Equivalent Rate of Neutrons

Calculations of the response for the most widely used neutron dosimeters at the Russian nuclear power plant (NPP) have been performed. It is shown that in some cases it is necessary to introduce a correction for the measured value of the ambient dose equivalent rate (AEDR). The experimentally tested values of the correction for measuring AEDR in the containment rooms of NPP with VVER-1200 are given.

Use of effective dose to assess x-ray protective clothing

This review article provides an overview on the results of studies conducted by the authors to improve the current personal protection concept in the clinical application of x-rays. With the aid of personal dose equivalent measurements during radiologically guided clinical interventions, laboratory tests using the Alderson-Rando phantom as well as Monte Carlo simulations various x-ray application scenarios were investigated. The organ doses and the effective doses of staff persons standing near the patient were determined. The 3D-attenuation properties of protective clothing under the scattered radiation emitted by the patient play a special role here. With regard to the minimisation of the quantity ‘effective dose’ the protection of the lower body from the gonads to the chest is of particular importance, since 80% of the effective dose is contributed by this region of the body. In contrast, protection of the back plays a subordinate role. Protective aprons optimised in terms of effective dose can be significantly lighter than conventional aprons, providing equal protection. The assessment of the attenuation properties of protective clothing should be based on the risk-related dose quantity, effective dose, rather than lead equivalent. In the future, the evaluation of radiation protective clothing could be based on the calculation of the effective dose assuming standardised irradiation conditions.

Measurement of Neutron Dose Equivalent within and Outside of a LINAC Treatment Vault Using a Neutron Survey Meter

This work concerns neutron doses associated with the use of a Siemens Primus M5497 electron accelerator, which is operated in the photon mode at 15 MV. The conditions offer a situation within which a fraction of the bremsstrahlung emission energies exceed the photoneutron threshold. For different field sizes, an investigation has been made of neutron dose equivalent values at various measurement locations, including: (i) At the treatment table, at a source-surface distance of 100 cm; (ii) at the level of the floor directly adjacent to the treatment table; and (iii) in the control room and patient waiting area. The evaluated neutron dose equivalent was found to range from 0.0001 to 8.6 mSv/h, notably with the greatest value at the level of the floor directly adjacent to the treatment couch (8.6 mSv/h) exceeding the greatest value on the treatment table (5.5 mSv/h). Low values ranging from unobservable to between 0.0001 to 0.0002 mSv/h neutron dose were recorded around the control room and patient waiting area. For measurements on the floor, the study showed the dose equivalent to be greatest with the jaws closed. These data, most particularly concerning neutron distribution within the treatment room, are of great importance in making steps towards improving patient safety via the provision of protective measures.

A Comparison of Conventional Empirical Formula and MCNPX Code in the Estimations of Photon and Neutron Skyshine Rates for an 18MV Radiotherapy Bunker

Purpose: A physical phenomenon, scattering the radiation by the atmosphere above the room to the points at ground level around the linac treatment room is known as skyshine radiation. This study aimed to estimate photon and neutron skyshine from a linac in a high-energy radiation therapy facility. Materials and Methods: The empirical method of NCRP report 151 and MC simulations were employed to estimate skyshine radiation dose from the 18MV linac photon beam. A linac and its bunker were modeled and skyshine dose equivalent from photons and secondary neutrons were derived and compared in the control room, corridor, sidewalk and, parking. Results: The photon skyshine dose rates calculations by the MC method varied from 0.43 µSv/h at the sidewalk to 6.2 µSv/h at the control room. The ratios of NCRP to MCNP calculations varied from 3.58 for the corridor to 16.14 for the control room. For the neutron skyshine dose rate at distances shorter than 20m, it was found to be 10.4 nSv/h and the ratios of the NCRP to MCNP were 1.26 at the control room and 3.34 at the sidewalk. Conclusion: It was concluded that the empirical method overestimates photon and neutron skyshine dose rates in comparison to the MCNPX code. The refinement of the proposed empirical method of NCRP 151 and application of MC methods are strongly suggested for more reliable calculations of skyshine radiations.