An HPGe detector-based shadow shield bed whole body counter has been developed for internal contamination monitoring of radiation workers. The system is calibrated for the measurement of internally deposited radionuclides which emits photon energies >100 keV. Its performance characteristics are compared with NaI(Tl) detector-based whole body counter. It is observed that due to superior energy resolution of the HPGe-based system, identification and quantification of internally deposited radionuclides in workers is superior than NaI(Tl) detector-based system. The ⁴⁰ K contents of 15 persons were measured using this system. It is found that the total body potassium varies from 0.82 gk kg ⁻¹ to 2.6 gk kg ⁻¹ of the body weight.

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The International Atomic Energy Agency (IAEA) coordinates and provides analytical support to the worldwide network of Analytical Laboratories for the Measurement of Environmental Radioactivity (ALMERA). The goal of ALMERA is to provide timely and reliable measurement results of environmental radioactivity in routine monitoring and emergency situations. Finding and choosing the most adequate analytical procedures for environmental monitoring can be a challenge for laboratories due to a wide variety of technologies available and the rapid developments in this field. To respond to this challenge, the IAEA has included within its ALMERA activities, the development and validation of a set of procedures for the determination of natural and anthropogenic radionuclides in environmental samples, for both routine and emergency environmental monitoring. Methodological harmonization, in turn, leads to enhanced worldwide comparability of environmental radioactivity measurement results. The respective analytical procedures are developed by expert groups from ALMERA member laboratories and validated through several ALMERA laboratories worldwide according to a comprehensive methodology. In addition to sequential analytical procedures for routine environmental monitoring, rapid procedures applicable for radiological emergencies have been developed and validated for the determination of americium and plutonium radioisotopes in soil and sediment samples, radiostrontium in milk, and radium isotopes in drinking water. Currently, the development of rapid procedures for radiostrontium analysis in soil and seawater is underway, partly prompted by the interest for high throughput procedures following the analytical burden of laboratories after the Fukushima Daiichi NPP accident.

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In view of an underestimation of personal dose equivalent due to ⁶⁰ Co radiation when system is calibrated to ¹³⁷ Cs, the effect of the atomic number of the filters used in the badge on the photon beams of 660 keV and 1250 keV was investigated. The low Z-high Z interface created between filters and discs was found to lead to discontinuity in dose at the surface of the disc. This effect was consistently observable due to the 0.8 mm thickness of the disc. It was concluded that under-estimation of dose due to ⁶⁰ Co radiation could be avoided by either reducing the thickness of the dosimeter disc or by a slight modification in the design of the filters.

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Effective atomic number for photon energy absorption (Z_peaeff), the relative effective atomic number of the thermoluminescent dosimetric (TLD) materials with respect to air (Z_reff), and energy dependence (ED) of the TLD materials such as beryllium oxide (BeO), lithium tetraborate (Li₂B₄O₇: Cu, Ag, P), aluminum oxide (Al₂O₃), calcium fluoride (CaF₂), and titanium oxide (TiO₂) in the energy range of 1 keV to 20 MeV have been studied. The values of ZPEA_eff and ZR_eff show a broad peak and a maximum value around 20 keV for BeO, around 60 keV for CaF2 and TiO₂, and around 40 keV for Al₂O3 and Li₂B₄O₇: Cu, Ag, P. Overall, the effective atomic number (Z_eff) varies linearly with ED for all the materials in the photon energy range of 1 keV to 20 MeV except TiO2 and thereby confirming the validity of a more commonly employed method of using the Z_eff to calculate the measure of the ED in TLD. The ED values show a broad peak around 30 keV for CaF₂ and TiO₂ and around 20 keV for Al₂O₃ and remain almost constant for the other two materials.

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For any civil engineering work, knowledge of extreme weather conditions at the site of interest is essentially required to design engineering structure that can withstand such extreme weather stresses. This paper presents an analysis of extreme values of meteorological parameters like temperature, estimated wind gust at 10 m height, wind speed, daily, monthly, and annual rainfall for 2004–2014. The yearly variations of these meteorological parameters were subjected to distribution analysis. Various distribution function parameters for each variable are determined. Extreme values corresponding to return periods of 50 and 100 years are worked out. These studies carried out in Kudankulam provide an insight into the extreme values of the considered variables over a period, generally during the expected lifetime of the reactor units and help the designer to arrive at the design basis values of different parameters as regards the safety of the units.

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Gamma-ray surveys are used in geological, geochemical, and environmental mapping for mineral exploration. Natural radioactivity levels for forty phosphate sample localities on the Eastern Desert, Nile Valley, and Western Desert, Egypt, were measured using a gamma-ray spectrometer with high accuracy. From the obtained results, the average values of the annual effective dose were 0.19, 0.28, 0.29, and 0.15 mSv/year for Safaga, Abu.Tartur, El-Sebaeya, and Hamrawayn, respectively. The measured values of natural radionuclides in the studied samples are higher than the world average activity values. This study could be useful as baseline data for radiation exposure to phosphate and their impact on human health.

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A new field-deployable technique for estimating the neutron ambient dose equivalent H*(10) by using the measured prompt gamma intensities emitted from borated high-density polyethylene (BHDPE) and the combination of normal HDPE and BHDPE with different configurations have been evaluated in this work. Monte Carlo simulations using the FLUKA code has been employed to calculate the responses from the prompt gammas emitted due to the monoenergetic neutrons interacting with boron, hydrogen, and carbon nuclei. A suitable linear combination of these prompt gamma responses (dose conversion coefficient [DCC]-estimated) is generated to approximate the International Commission on Radiological Protection provided DCC using the cross-entropy minimization technique. In addition, the shape and configurations of the HDPE and BHDPE combined system are optimized using the FLUKA code simulation results. The proposed method is validated experimentally, as well as theoretically, using different workplace neutron spectra with a satisfactory outcome.

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The occupational workers of fuel fabrication and reprocessing facilities at Indira Gandhi Center for Atomic Research (IGCAR) have a potential for internal exposure to natural uranium which is hazardous both in chemical and radiological aspects. Hence, in vivo monitoring of the radiation workers has to be carried out to ensure safe working conditions. In IGCAR, the in vivo monitoring of natural uranium is being carried out using Phoswich-based lung monitor. The measurement and quantification of internal exposure due to natural uranium is done using 63 and 93 keV photons emitted by ²³⁴ Th, immediate daughter of ²³⁸ U. Realistic anthropomorphic Lawrence Livermore National Laboratory (LLNL) phantom is used for the calibration of the system. As natural uranium-loaded lung set is not available currently in the laboratory, lung set loaded with indigenously prepared natural uranium source capsules were used. Efficiency curve for ²³⁸ U was established for varying muscle equivalent-chest wall thickness (MEQ-CWT) and the efficiency values were in the range of 5.604E-03 to 8.601E-03 CPS/Bq. Simulation results of LLNL voxel phantom having uniform lung distribution of natural uranium agreed within 9% with the measured one which is comparable with the error associated with measurement. This confirms that the distribution pattern of 12 capsules in each lung in the given geometry closely resembles the uniform distribution. The efficiency and minimum detectable activity (MDA) values for Indian population were found to be ranging from 1.011E-03 to 7.931E-03 CPS/Bq and 12–18 Bq, respectively. The efficiency value for ²³⁸ U established from the measurement using Japan Atomic Energy Research Institute (JAERI) phantom having uniform source distribution agreed with that of LLNL phantom measurement having capsule source distribution for the same MEQ-CWT thickness (33.8 mm) within 3%. This further reaffirms that the adopted capsule distribution is close to uniform distribution.

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Studies have been carried out to assess the radiological exposure to the general public from small-scale mining activities in Dunkwa-on-Offin and its surrounding communities. Direct gamma spectrometry was used to determine the concentration of naturally occurring radionuclides ²²⁶ Ra (²³⁸ U),²³² Th, and ⁴⁰ K in the soil samples. The mean activity concentrations measured for ²²⁶ Ra (²³⁸ U),²³² Th, and ⁴⁰ K in the soil samples were 25.4 ± 11.1, 29.4 ± 15.6, and 225.9 ± 93.8 Bq/kg, respectively. The total annual effective dose to the public was estimated to be 0.05 mSv. The results, thus, indicate an insignificant exposure of the general public. The radiological parameters assessment as a result of ²²⁶ Ra (²³⁸ U),²³² Th, and ⁴⁰ K was also carried out. Data established from the study could aid in formulating guidelines, to educate and create awareness on the levels of naturally occurring radionuclides resulting from small-scale mining activities in Dunkwa-on-Offin and Ghana as a whole.

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