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.

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.

X-ray Emission Hazards from Ultrashort Pulsed Laser Material Processing in an Industrial Setting

Interactions between ultrashort laser pulses with intensities larger than 1013 W/cm2 and solids during material processing can lead to the emission of X-rays with photon energies above 5 keV, causing radiation hazards to operators. A framework for inspecting X-ray emission hazards during laser material processing has yet to be developed. One requirement for conducting radiation protection inspections is using a reference scenario, i.e., laser settings and process parameters that will lead to an almost constant and high level of X-ray emissions. To study the feasibility of setting up a reference scenario in practice, ambient dose rates and photon energies were measured using traceable measurement equipment in an industrial setting at SCHOTT AG. Ultrashort pulsed (USP) lasers with a maximum average power of 220 W provided the opportunity to measure X-ray emissions at laser peak intensities of up to 3.3 × 1015 W/cm2 at pulse durations of ~1 ps. The results indicate that increasing the laser peak intensity is insufficient to generate high dose rates. The investigations were affected by various constraints which prevented measuring high ambient dose rates. In this work, a list of issues which may be encountered when performing measurements at USP-laser machines in industrial settings is identified.

Measurement of ambient dose rates in tantalite mining sites in Oke ̶ Ogun, southwest, Nigeria

The work scenarios involved in the mining of tantalite a radioactive material expose the miners to ionizing radiation from the ore and the surrounding environment. The dose level in the mine air may be higher than the safe limit due to various contributory sources of ionizing radiation such as radionuclides from rocks, effluents, sand, and radon gas that emanates from caves and this can be of health detriment to the miners. Measurements of ambient dose rates in four selected mining sites have been investigated. Gamma absorbed dose rates were measured in air onsite at Komu, Sepenteri, Gbedu, and Eluku mining sites in Oke-Ogun areas of Oyo State, Nigeria using GammaRAE II dosimeter. Radiation dose to risk software was used to estimate the cancer risk for the period the miners spent onsite. The measured mean dose rate at the sites falls within the range of (19-240) nSv/y and the estimated annual dose rate, cumulative dose, and cancer risk fall within the range of (37-314) μSv/y, (4.0 ̶ 11.1) mSv and (0.5 ̶ 4.5) E-04 respectively. The upper limits of the range for the radiological parameters are all above the safe limit. The health implication of that is that increased work activities at these mining sites may over the years have a negative health effect on the miners. The exposure time of workers can be reduced through proper planning of working shifts for the miners.

Development of neutron survey meter using prescila neutron probe

This paper presents the design and validation of a neutron survey meter. The meter consists of a PRESCILA neutron probe (with good sensitivity, directional response, gamma rejection, and enhanced high-energy response to 20 MeV) and an electrometer developed at Non-Destructive Evaluation center. The homogeneity response of the PRESCILA neutron probe was investigated as a function of distances from the 241Am - 9Be source in order to obtain the appropriate distance for accurate count-rate measurements using the neutron survey meter. A system consists of the PRESCILA neutron probe and the Ludlum Model 2326 electrometer was then used for measuring neutron ambient dose equivalent rates in the range from 50 cm to 200 cm with the step of 25 cm. The relationship between the count-rates and neutron dose equivalent rates (in the distance ranged from 50 to 200 cm) were deduced to validate the proper operation of the neutron survey meter.

Characterization of an ²⁴¹Am-Be neutron irradiation facility at Institute for Nuclear Science and Technology

An automated panoramic irradiator with a 241Am-Be neutron source of 5 Ci is installed in a bunker-type medium room at the Institute for Nuclear Science and Technology (INST) for calibration of neutron devices. Bonner Sphere Spectrometer (BSS) formed by 6 spheres plus bare detector, with cylindrical, almost point like, 6LiI(Eu) scintillator and 2 different spectral unfolding FRUIT and BUNKIUT codes are used to characterize the neutron field in different measurement points along the irradiation bench. The neutron field is also simulated by MCNP5 software and compared with measurements performed by the BSS. The paper shows the main results obtained in terms of neutron spectra at fixed distances from the source as well as their neutron fluence rate (totaland direct) and ambient dose equivalent rate. These values measured by the BSS with two unfolding FRUIT and BUNKIUT codes are in good agreement with that of simulated by MCNP5 within 10%.

Development of a neutron detector for radiation protection monitoring

The paper presents the results of the development of a neutron detector for radiation protection purposes. Monte Carlo simulations, using MCNP5 code, were performed to optimize the configuration of the neutron detector. The developed detector consists of a 3He proportional counter embedded in a multi-layer moderator made of high-density polyethylene (HDPE) and Cadmium. The characteristics of the developed neutron detector including neutron fluence response and ambient dose equivalent response were calculated, analyzed and compared with those from other neutron survey meters. The simulation model and computed results were assessed through experimental measurements at the Secondary Standards Dosimetry Laboratory of the Institute for Nuclear Science and Technology (INST). A good agreement between the simulated and experimental results was observed within 9.3% for 241Am-Be source and four simulated workplace neutron fields.