Barite concrete-based cement composites for ²⁵²Cf spontaneous neutron and ⁶⁰Co/¹⁹²Ir shielding based on Monte Carlo computation

Abstract Barite concrete composite materials have been investigated for 252Cf spontaneous neutron and 60Co/192Ir gamma sources’ shielding using Monte Carlo computational method. The Particle and Heavy Ion Transport code System (PHITS) was used to compute the shielding properties of three different materials (barite concrete, barite cement, and barite aggregate) used as structural walls in fixed neutron & gamma industrial radiography for Non-Destructive Testing applications. The obtained results displayed good properties of barite concrete in shielding spontaneous neutrons emitted from the 252Cf source, as the effective dose drops about 108 times in only 140 cm wall thickness, and it was found to be about 10 times more effective than other materials investigated. In addition, the investigated gamma shielding properties of the barite concrete showed a relatively smaller wall thickness compared to the ordinary concrete. The decision-making process based on the ALARA principle of dose limitation showed that the use of barite concrete in such facilities is more effective than the use of barite cement and barite aggregate, for both gamma and neutron radiography shielding design. To achieve an average value of 1 μSv/h, the obtained result shows that 80 cm of Barite concrete is needed, while 125 and 130 cm of barite cement and barite aggregate are needed, respectively to shield the Co-60 source. Meanwhile, 50 cm of wall made of barite concrete is sufficient to cut down the effective dose rate to 1 μSv/h (for 50 Ci and 55 cm for 150 Ci 192Ir), which is an appropriate design for the public area adjacent to the industrial radiographic facility. It was therefore concluded from the obtained data that barite concrete is the most effective shielding material for radioactive sources (60Co, 192Ir, and 252Cf) used in radiographic applications

Precision measurement of radioactivity in Gamma-rays spectrometry using two HPGe detectors (BEGe-6530 and GC0818-7600SL models) comparison techniques: Application to the soil measurement.

To obtain high quality of results in gamma spectrometry, it is necessary to select the best HPGe detector for particular measurements, to calibrate energy and efficiency of gamma detector as accurate as possible. To achieve this aim, the convenient detector model and gamma source can be very useful. The purpose of this study was to evaluate the soil specific activity using two HPGe model (BEGe-6530 and GC0818-7600SL) by comparing the results of the two detectors and the technics used according to the detector type. The relative uncertainty activity concentration was calculated for 226Ra, 232Th and 40K. For broad energy germanium detector, BEGe-6530, the relative uncertainty concentration ranged from 2.85 to 3.09% with a mean of 2.99% for 226Ra, from 2.29 to 2.49% with a means of 2.36% for 232Th and from 3.47 to 22.37% with a mean of 12.52% for 40K. For GC0818-7600SL detector, it was ranged from 10.45 to 25.55% with a mean of 17.10% for 226Ra, from 2.54 to 3.56% with a means of 3.10% for 232Th and from 3.42 to 7.65% with a mean of 5.58% for 40K. The average report between GC0818-7600SL model and BEGe-6530 model was calculated and showed the mean value of 3.36. The main study was based on the following points: • Determination of The relative uncertainty activity concentration of 226Ra, 232Th and 40K • Determination of the relative uncertainty related to the radium equivalent activity to compare the performance of the two detection systems • Proved that the activity concentration determination in gamma spectrometry depended on the energy range emitted by a radionuclide. This study showed that the standard deviation measurement was less important to the result realized with BEGe-6530 HPGe model. Our findings were demonstrated that the results of the Broad Energy Germanium detector were more reliable

Optimal measurement counting time and statistics in gamma spectrometry analysis: The time balance

The optimal measurement counting time for gamma-ray spectrometry analysis using HPGe detectors was determined in our laboratory by comparing twelve hours measurement counting time at day and twelve hours measurement counting time at night. The day spectrum does not fully cover the night spectrum for the same sample. It is observed that the perturbation come to the sun-light. After several investigations became clearer: to remove all effects of radiation from outside (earth, the sun, and universe) our system, it is necessary to measure the background for 24, 48 or 72 hours. In the same way, the samples have to be measured for 24, 48 or 72 hours to be safe to be purified the measurement (equality of day and night measurement). It is also possible to not use the background of the winter in summer. Depend on to the energy of radionuclide we seek, it is clear that the most important steps of a gamma spectrometry measurement are the preparation of the sample and the calibration of the detector

Assessment of natural radioactivity and associated radiation hazards in sand building material used in Douala Littoral Region of Cameroon, using gamma spectrometry

Twenty-four sand samples were collected from different sand quarries from Douala Littoral Region (Wouri, Dibamba, Mungo and Docteur Anse rivers and Atlantic Sea) along the Guinea Golf. These samples were investigated using gamma-ray spectrometry system. Highest values of 226Ra, 232Th and 40K measured specific activities expressed in Bq Kg−1 units were, respectively, 146.7 (in Youpoue–Bamenda 2) 102.9 (in Village 1) and 928 (in Northern Akwa 6) while the lowest values were found to be, respectively, 11.8 (in Northern Akwa 6), 8.0 (in Bonaberi–Bonamikano 4) and 54.0 (Youpoue 3). The potential radiological hazards parameters were assessed by calculating successively radium equivalent activity (Raeq), outdoor absorbed gamma dose rate (Dout), annual effective dose rate, internal hazard (Hin) and external hazard (Hex) indices and alpha and gamma index from using those sand in the construction of dwellings and large buildings. Results obtained show that annual dose absorbed by inhabitants due to sand construction use in Douala is below 1.0 mSv year−1. Therefore, most of the types of sands studied and incorporated in constructions appear to be safe as building material. The outputs from this research will be useful to assess the radiation hazards of sand building material in humans and to initiate a sand database together with a radiological map of the area at stake