An important parameter of Geiger Muller Detectors is their lengthy dead time which causes a nonlinear response at high counting rates of the Detector. Dead time depends on the Detector’s nature (size, geometry and material) and its connected electronics (detection, counting, and quenching circuits) plus detection settings. A very important and efficient method of dead time cancelation is time to first count method of measurement at the expense of a complicated electronic circuitry for active quenching of the Geiger Muller Detector. In this paper, based on FPGA, a fast and efficient active quenching circuit is developed. Moreover, the time to first count (named as time intervals in this paper) is also acquired at the same time by the system on a chip (SoC) architecture. The time resolution of the system is 10 ns which is pretty enough for Geiger Muller Detectors regarding their timing characteristics (the dead time of the Geiger Muller Detectors ranges from several microseconds to more than 200 microseconds). ZP1210 Geiger Muller Detector is taken as a case study. The dead time of the Detector is about 200 µs according to the manufacturer datasheet. This active quenching circuit which is implemented on a single chip, is experimentally tested on this Detector. Details of the work is explained in the text.

An in-depth fuel assay is crucial for spent nuclear fuel management. Spent fuels with high radiation dose levels also require an appropriate dosimeter. A widely used Detector for nuclear radiation dosimetry is the Geiger-Muller Detector. Pulses from this Detector are of the same amplitude and no information about particle energy is provided. Dead time is a distorting effect as a nonlinear response is observed at large counting rates. This requires correction methods to avoid nonlinearity. The special measuring tool must be consistent with the structures and constraints of spent nuclear fuels, spent fuel storage, and its construction design. Spent nuclear fuel assemblies are kept in 10-meter water in the open reactor pool for a specific cooling period before transportation to the spent fuel storage pool. Measurement of high dose rates requires specific equipment. In this work, a spent fuel active dosimeter is developed for the spent fuel assay in TRR. Due to high radiation exposure, equipment parts are specially designed to resist gamma radiation. To eliminate the pulse pile-up effect and noise cancellation, a bandpass filter is employed. As an advanced technique, time interval distribution (TID) is also developed using digital electronics. By utilizing a Co-60 standard gamma source in the Karaj Secondary Standard Dosimetry Laboratory (SSDL), the Detector tests and calibration are also accomplished. Validation of the system is performed with a commercial measuring tool.

In this paper we consider a gaseous Detector and supposed, because of pass of an ionizing radiation, an electron created inside it. By numerical simulation with monte carlo method and concluding the impacts, scatterings and creation of secondary electrons, we find the trajectory of initial and secondary electrons. Dependence of number of secondary electrons to applied electrical field is investigated. Finally we study the dependence of number of secondary electrons to number of initial electrons created by ionizing radiation (as a measure of the radiation energy), then we try to investigate the possibility of spectrometry by this Detectors.

Background and Objective: Due to importance of ionizing radiation on human health, many studies have been performed to measure the background gamma radiation all around the world as well as some cities in lran. This study was carried out to measure the amount of background gamma radiation in outdoor areas in different seasons in Zanjan to determine the annual effective dose of the city residents.Materials and Methods: To determine the dose rate of background gamma radiation in outdoor areas, 8 stations (4 in the main directions and 4 in the downtown areas) were selected using the map of the city. Eight measurements were performed for each station (twice in each season) using Geiger-Muller Detector (RDS-110) calibrated by Cs-137.Results: The mean value of dose rate and the annual effective dose due to background gamma radiation in different season in Zanjan were determined 126 nGy/h and 0.15 mSv respectively. The minimum and maximum mean values of dose rate were found 120±21 nGy/h and 134±18 nGy/h in summer and spring respectively.Conclusion: The results show that the dose rate and the annual effective dose for the city residents due to the background gamma radiation in outdoor areas is twice as much as international mean value reported by UNSCEAR-2000. To determine the total annual effective dose of Zanjan residents, measuring the dose rate in indoor areas is necessary.

This paper has investigated the effect of operating parameters such as ambient temperature and applied voltage on the dead time of a thin-walled Geiger-Muller (GM) counter using non-paralyzing model and two-source method. Experimental studies have been conducted using 137Cs and 90Sr sources at voltages ranging from 600 to 800 V and in the temperature range of-27-70 0C. The results of the investigations for applied voltage indicate that the dead time behavior in terms of voltage can be classified into three distinct regions. In region I (low voltages), the dead time decreases with increasing voltage, in region II (voltage close to the operating voltage), the dead time is almost constant. The dead time in region III (voltages above 740V) increases slowly and linearly with increasing voltage. The variation in the dead time in the region I is greater than region III. Region II with the minimum dead time and minimum variation of the operating voltage is the best operating region. Studies show that the variation of dead time and the range of the dead time plateau (District II) for 137Cs and 90Sr sources is different. The dead time was using 137Cs source was obtained between 58 and 78 ms and using 90Sr source between 79 and 130 ms. In general, the variations of dead time versus voltage for each of Regions I, II and III for 90Sr source are lower than 137Cs source. Experimental results also show that the dead time increases with increasing temperature.

Backgrounds and Objectives: Due to importance of ionizing radiation on human health, many studies have been performed to measure the background gamma radiation all around the world as well as some cities in lran. According to this fact that everybody spends almost 80% of his time in indoor areas, this study was carried out to measure the amount of background gamma radiation in indoor areas in Zanjan city located in northwest of Iran to determine the annual effective dose of the city residents.Materials and Methods: To determine the dose rate of background gamma radiation in indoors, 30 dwelling (in the main geographical directions and in the downtown areas) were selected. All of them were one floor and iron roofed. A Geiger-Muller Detector (RDS-110) calibrated by Cs-137 was used in each livings room of each dwellings that performed in one meter far from the earth. In 30 minute 30 values was recorded. The mean value was considered as indoor gamma dose rate in each dwelling.Results: The mean value of dose rate in Zanajn indoor areas due to gamma background radiation was determined 146±25 nGy/h. According to the results and findings in our previous study about gamma background radiation in outdoor areas in Zanjan, the annual effective dose was determined and 0.87 mSv for Zanjan city residents.Conclusion: The annual effective dose of Zanjan city residents due to the gamma background radiation is 0.87 mSv that in comparison with UNSCEAR-2000 report is higher than the mean value for the world (0.48 mSv).

This study aimed to produce a personal dosimeter to connect the electronic devices to transfer the dosimetry data online including dose rate data and integral dose. At first, an electronic board was designed. Afterwards, the board was simulated using Altium Designer software. Then, the microcontroller programming was performed using Code Vision software. The performance of the elements was tested using Proteus Design Suite software and the circuit diagnosis and testing were practically implemented. Afterwards, the calibration factor and the count-to-dose conversion coefficient were obtained using an ionization chamber in the Secondary Standard Dosimetry Laboratory of Iran. Furthermore, dose distribution as a function of distance to the source, was obtained using the produced dosimeter and was compared using a calibrated ionization chamber. To this purpose, an unmitigated Cs-137 source, a Cs-137 source with a 22 mm Pb shield, and a Cs-137 source with a 22 mm plus 18 mm shield were used. The results showed a calibration factor of 0. 99 ± 0. 02 for the designed dosimeter. Calibration results showed that the produced dosimeter passed the ISO/IEC 17025: 2005 criteria. Furthermore, the dose distribution results demonstrated negligible differences of less than 0. 01% between the produced dosimeter and the reference ionization chamber. It was concluded that the produced dosimeter could be used as a reference and standard dosimeter in laboratories, for monitoring and dosimetry of ionizing photon radiations.

Background: The annual effective dose equivalent (AEDE) due to ionizing radiation in a coaster area in Nigerian has been determined using Geiger-Muller counter. Materials and Methods: isolevels in the study area were represented by using the kriging interpolation technique on ArcGIS 10. 1 software. Duncan multiple range test was performed using Statistical Package for Social Sciences (SPSS). Results: The results obtained show that the AEDE ranged from 0. 19± 0. 01 to 0. 35± 0. 02 mSv y-1 with a mean of 0. 27± 0. 03 mSv y-1. The excess lifetime cancer risk (ELCR) ranged from 0. 0007 to 0. 0012, with a mean of 0. 0010. The total annual collective effective dose equivalent is 4839. 49 person-Sv. Conclusion: The AEDE in the study area is lower than the maximum permissible limit of 1 mSv y-1 recommended by ICRP. The AEDE was higher in areas of high elevations than those of low elevations. It was estimated that about twenty-eight (28) of every one hundred thousand (100, 000) persons are at risk of developing radiation-induced diseases per annum in the study area.

Background: Measurement of background radiation is very important from different points of view especially for human health. The aim of this survey was focused on determining the current background radiation in one of the highest altitude regions ( Zagros Mountains ), Chaharmahal and Bakhtiari province, in the south west of Iran . Materials and Methods: The outdoors-environmental monitoring exposure rate of radiation was measured in 200 randomly chosen regions using portable Geiger-Muller and Scintillation Detectors. Eight measurements were made for each region and an average value was used to calculate the exposure rate from natural background radiation. Results: The exposure dose rate was found to be 28.4 m Rh-1 and the annual average effective equivalent dose was found to be 0.49 mSv. An overall population weighted average outdoor dose rate was calculated to be 49 nGyh-1, which is higher than the world-wide mean value of 44 nGyh-1 and is comparable to the annual effective equivalent dose of 0.38 mSv. Conclusion: A good correlation between the altitude and the exposure rate was observed, as the higher altitude regions have higher natural background radiation levels

The entire globe is radioactive naturally, and humans are constantly exposed to background radiation from cosmic rays and the radioactive materials in their environment. The concentration and effects of background radiation can vary based on geographical location. Measuring background radiation levels is important for assessing potential health impacts. This study presents a comprehensive data analysis to investigate the levels and impact of background radiation levels in Sahiwal, Pakistan, and determine if the levels are safe according to international standards. Radiation counts were measured using a Geiger-Muller counter at several locations in Sahiwal over 40 days. The data was analyzed using normal distribution techniques to calculate the effective absorbed dose of the ionizing radiation in human tissue. The calculated dose was then compared to internationally accepted safe exposure levels. The effective absorbed dose of ionizing radiation in Sahiwal was determined as 0.27 mSv/year, significantly lower than the worldwide average background dose of 2.4 mSv/year. Based on this result and comparisons to international standards, the study concluded that Sahiwal is a safe area in terms of background radiation exposure for human living. However, more comprehensive measurements over longer periods could provide additional insights.