Microstructural Investigation of PM-355 Nuclear Track Detector Subjected to Low-Dose Gamma Irradiation: A Positron Annihilation Lifetime Study

Samples of the PM-355 polymeric solid state nuclear track detector were exposed to low gamma absorbed doses from 1 kGy (0.1 Mrad) up to 9 kGy (0.9 Mrad). Positron annihilation lifetime (PAL) in conjunction with transmission electron microscopy (TEM) and Thermo-gravimetric analysis (TGA) were performed on irradiated and pristine samples at room temperature. The observed lifetime spectra were resolved into three components where the ortho-positronium (o-Ps) lifetime component was associated with the pick-off annihilation of positronium trapped by the free volume. PAL studies of irradiated PM-355 samples showed that ortho-positronium (o-Ps) lifetime increases with an increase in dose up to 4 kGy and decreases at higher doses. In contrast, the intensity of the o-Ps component, I3, decreases with the dose up to about 2 kGy, followed by a much smooth decrease up to 7 kGy, and then it levels off. TGA analysis indicated that the PM-355 detector decomposed in one main breakdown stage. These results are discussed on the basis of chemical and physical changes occurring at the microscopic level in the PM-355 due to irradiation. Crosslinking dominates for doses between 1 and 4 kGy, while the degradation mechanism (Chain scission) prevails for doses up to 9 kGy.

Analytical and statistical calculation of gamma dose rate for the accident of losing the shield for Tehran Research Reactor

In this paper we study the analytical and statistical results of estimating the gamma dose rate at pool access floor in TRR when the core shield accidentally decreases to some non-permitted levels. Due to the risk of experimental techniques, we use the analytical and statistical methods. In normal conditions （no risk）, the discrepancies between experiment and two methods are justified and it is found that for such problems we have to normalize these methods to experimental results as follows： the analytical method by factor 0.13 and MCNP by 1.7.

Dose inter-comparison studies for ^(60)Co gamma-ray and electron beam irradiation in the year 2002

Dose inter-comparison studies for Co γ-ray and 10 MeV electron beam irradiation were carried out 60 from July to October in 2002. The purpose of the studies was to check the reliability of the alanine-PE film dosime- ters made by CIAE, which will be used as transfer standard dosimetry system mainly for electron beam irradiation. The expanded uncertainty of CIAE alanine/EPR dosimetry system was 4.1% for doses not higher than 10 kGy and 5.4% for those above 10 kGy (k=2). CIAE alanine-PE film dosimeters were sent to JAERI, RISO (National Labora- tory in Denmark) and INCT respectively, which were irradiated by Co gamma-rays or electron beams in each labo- 60 ratory. The irradiated dosimeters were then sent back to CIAE for electron paramagnetic resonance (EPR) analysis. The agreements were obtained to be ±1.9% for gamma-ray dose measurement and ±4.3% for electron beam dose measurement, which were all within the combined uncertainty of the reference and CIAE alanine/EPR dosimetry system. Furthermore, the overall mean ratio was found to be 0.995 with 1.8% in the coefficient of variation (CV). The preliminary inter-comparison studies indicated that CIAE film alanine/EPR dosimetry system had the potential to be used as a transfer dosimetry system for high dose measurement.

Dose Rate Calculation in the Vicinity of the Tunisian Gamma Irradiation

Dose rates calculations, in PMMA dosimeters, placed in the vicinity of the Tunisian 60Co gamma ray irradiator, have been achieved using a pencil like model. The obtained results are in good agreement with recent experimental data. Moreover, in this work we determine also the conversion factor between the dose rate deposited in a PMMA dosimeter and the one deposited in a reel medium. This factor is used to determine the dose deposited in a real irradiated medium such as foodstuff products.

Comparison of Gamma Pass Rate between the Dose-to-Water and Dose-to-Medium Reporting Modes for Patient-Specific QA Using a Helical Diode Array Dosimeter with a Fixed Phantom Density

Introduction: To compare the measured dose distributions to calculated ones in dose-to-water (Dw) and dose-to-medium (Dm) reporting modes for simple plans and patient-specific intensity modulated radiation therapy (IMRT) and volumetric modulated arc therapy (VMAT) plans using ArcCHECK with a fixed phantom density. Methods: The recommended density value of 1.18 g/cm3 for Acuros XB and X-ray voxel Monte Carlo was assigned to ArcCHECK on CT images. A total of 45 simple plans, including a 1-field plan, a 3-field plan, a 4-field plan, a half-arc plan from 270° to 90°, and a full-arc plan, were assessed. Subsequently, the patient-specific 96 IMRT and VMAT plans were evaluated. Gamma analysis with a 3% normalized global dose error and a 3 mm distance-to-agreement criteria (γ3%G/3mm) was performed in the Dw and Dm. The change in γ3%G/3mm between Dw and Dm were statistically analyzed using JMPPro11 software. Results: The median values of γ3%G/3mm for all simple plans for Dw and Dm were 98.1% (range, 75.2% - 100%) and 95.5% (range, 23.7% - 100%), respectively (p 0.01). In the patient-specific IMRT and VMAT plans, the median values of γ3%G/3mm for Dw and Dm were 98.6% (range, 90.1% - 100%) and 90.5% (range, 38.5% - 97.2%), respectively (p 0.01). Conclusion: Our results showed that the calculated and measured dose distributions were in good agreement for Dw, but were not for Dm. From the viewpoint of the rationale of dosimetry, Dw shows better agreement with measured dose distribution when using the fixedphantom density recommended by the vendor.

Assessment of Equivalent Radium Activity and Annual Effective Dose Due to Building Materials in Côte D’Ivoire by Gamma Spectrometry: Cases of Cement

Radioactivity is a natural phenomenon present in the universe. So, because of human life solidarity with a habitat, we are permanently in contact, especially through building materials. The objective of this work is the determination of the used cement radioactivity level in the buildings in Cote d'Ivoire. Thus, samples of grey and white cement currently used on Ivorian territory were taken. In order to determine the radioactivity level of this cement, samples were analyzed by using gamma spectrometry chain which contains a NaI (Tl) scintillation detector designed by the German manufacturer LD-DIDACTIC, coupled to a multichannel analyzer (AMC) using a Cassy Lab software. Thus, the specific activity of the primordial radionuclides 226Ra, 232Th and 40K, was able to be determined. The average values obtained are 29.66 Bq/kg, 34.88 Bq/kg and 178.424 Bq/kg respectively for 226Ra, 232Th and 40K. All average values are below the limit values recommended by UNSCEAR. However, we evaluated the radiological parameters such as the equivalent radium activity and the annual effective dose in order to translate the specific activity in terms of harmfulness. Values obtained for these parameters are below those recommended by ICRP and UNSCEAR. These results show that the risk incurred by the use of these different brands of cement is low.

SRS MapCHECK与EDOSE在立体定向放疗计划剂量验证中的对比

目的:使用SRS MapCHECK和EDOSE对立体定向放疗(SRT)计划进行剂量验证,并比较它们的验证结果。方法:首先比较两种系统在不同方野(2 cm×2 cm、4 cm×4 cm、6 cm×6 cm、10 cm×10 cm)的γ通过率(3%/3 mm),然后选择29例SRT计划,分析两种系统在不同γ标准下的绝对剂量通过率,最后分析EDOSE建立的4种不同物理模型对SRT计划剂量验证的影响。结果:SRS MapCHECK在所有方野和29例SRT计划的绝对剂量γ通过率均高于EDOSE,两者的验证结果有统计学差异(P<0.05);SRS MapCHECK及EDOSE在29例SRT计划的γ通过率(2%/2 mm)分别是为98.60%±2.14%和96.53%±2.41%。基于EDOSE的不同物理模型验证结果有统计学差异,平均γ通过率(2%/2 mm)偏差为1.8%~5.1%。结论:SRS MapCHECK和EDOSE系统均满足SRT计划剂量验证的要求,且SRS MapCHECK的剂量验证通过率优于EDOSE;不同的EDOSE物理模型对SRT计划剂量验证有影响。

Monte Carlo simulation of single beam dose profile

The single beam dose profiles of Leksell Gamma Unit are simulatedand the results are compared with the experimental results of ELEKTA. The effect ofdiameter of cylinder source on the dose profile is also studied and the yielded resultsshow that there are not much difference of the dose profiles between a point sourceand cylinder source when diameter is less than 1mm.

Measurements of ^(134)Cs and ^(137)Cs in urine and estimation of the internal dose of an adult exposed to the Chernobyl Accident

To estimate the internal dose of a Chinese visiting scholar internally contaminated because of the Cherno- byl Accident, the contents of 134Cs and 137Cs in urine were measured using a Ge(Li) γ-spectrometer. The internal doses were calculated based on data from the ICRP Publications. The effective doses from 134Cs and 137Cs were estimated to be 61 μSv and 98 μ Sv, respectively. The sum of 159 μSv was lower than the total effective dose (310 μSv), from the inhalation and ingestion of natural radionuclides. The dose of 131I was also reviewed referring to the UNSCEAR 2000 Report. The equivalent effective dose of 131I was estimated to be 2.9 mSv, 18 times more than the amount of 134Cs and 137Cs. Therefore, it is considered that the earlier estimation of internal doses of 131I is important in evaluating radiation injuries from a nuclear reactor accident.

Effect of ionizing radiation on dual 8-bit analog-to-digital converters (AD9058) with various dose rates and bias conditions

The radiation effects on several properties （reference voltage, digital output logic voltage, and supply current） of dual 8-bit analog-to-digital （A/D） converters （AD9058） under various biased conditions are investigated in this paper. Gamma ray and 10-MeV proton irradiation are selected for a detailed evaluation and comparison. Based on the measurement results induced by the gamma ray with various dose rates, the devices exhibit enhanced low dose rate sensitivity （ELDRS） under zero and working bias conditions. Meanwhile, it is obvious that the ELDRS is more severe under the working bias condition than under the zero bias condition. The degradation of AD9058 does not display obvious ELDRS during 10-MeV proton irradiation with the selected flux.

Simulation of radiation dose distribution and thermal analysis for the bulk shielding of an optimized molten salt reactor

The Chinese Academy of Science has launched a thorium-based molten-salt reactor(TMSR)research project with a mission to research and develop a fission energy system of the fourth generation.The TMSR project intends to construct a liquid fuel molten-salt reactor(TMSR-LF),which uses fluoride salt as both the fuel and coolant,and a solid fuel molten-salt reactor(TMSR-SF),which uses fluoride salt as coolant and TRISO fuel.An optimized 2 MWth TMSR-LF has been designed to solve major technological challenges in the Th-U fuel cycle.Preliminary conceptual shielding design has also been performed to develop bulk shielding.In this study,the radiation dose and temperature distribution of the shielding bulk due to the core were simulated and analyzed by performing Monte Carlo simulations and computational fluid dynamics(CFD)analysis.The MCNP calculated dose rate and neutron and gamma spectra indicate that the total dose rate due to the core at the external surface of the concrete wall was 1.91μSv/h in the radial direction,1.16μSv/h above and 1.33μSv/h below the bulk shielding.All the radiation dose rates due to the core were below the design criteria.Thermal analysis results show that the temperature at the outermost surface of the bulk shielding was 333.86 K,which was below the required limit value.The results indicate that the designed bulk shielding satisfies the radiation shielding requirements for the 2 MWth TMSR-LF.

Measurement of Gamma-Rays Using Smartphones

The CMOS (Complementary Metal Oxide Semiconductor) image sensor of a smartphone has been known for its sensitivity to gamma-rays. In this research, some smartphones were selected and tested for measurement of gamma-rays emitted from Cesium-137 and Iridium-192 sources. During measurements, the phones were set in video mode while the camera lenses were covered with black adhesive tape to prevent light exposure. Interaction of gamma-rays with the CMOS appeared as flashing bright spots on the image. The bright spots were then counted by using the freely available ImageJ software. Preliminary results indicated that the number of bright spots increased linearly with increase of gamma-ray dose rate. An in-house Android application software was then developed for real-time counting of the bright spots. The application software also allowed users to input a calibration equation so that the phones could simultaneously convert the count rate to display in dose rate. This research demonstrated that, after appropriate calibration, smartphones could be used as gamma-ray measuring devices for radiation safety control involving high activity sources such as in industrial radiography, gamma-ray irradiation facility and medical treatment.

Investigation of Gamma-Emitting Natural Radioactive Contents in Three Types of Vernonia Consumed in Cameroon

The specific activity concentration and the derived Annual Effective Dose(AED) in three types of vernonia cultivated and consumed within and outside Cameroon were measured by means of a well-calibrated high-purity germanium detector. Vernonia samples were collected directly from the production farms, oven-dried to a constant mass, crushed, sieved and sealed for at least a month before analysis. The specific activity of 238U in the three types of vernonia ranged from 20 - 50 Bq kg-1 with an average of 42 ± 15 Bq kg-1, 232Th from 9 - 22 Bq kg-1 with an average of 17 ± 7 Bq kg-1 while 40K ranged from 115 - 460 Bq kg-1 with an average of 302 ± 36 Bq kg-1. The average AED for 40K, 238U and 232Th were 0.15, 0.92 and 0.92μSv y-1 respectively. 238U and 232Th show the same trends both for the regional distribution of the radioactivity content and the AED. The discrepancies in our data can be attributed to many factors such as geological formation, foliar deposition, type and age of the plant, etc. Although the results obtained represent only some fractions of the standard limit, but they are within some range obtained in other countries.

Gamma radiation levels in the ambient environment of the QNPP Base

Monitoring results of gamma dose rate level in 1992～2004 in the ambient environment of the Qinshan Nuclear Power Plants(QNPP)Base,the northeast of Zhejiang Province,are reported in this paper.It is shown that the gamma dose rate of five monitoring sites of 2.5 km to QNPP Base is 84～113 nGy/h,with an average of 96 nGy/h in the 13 years.The average value is close to the background level of 93 nGy/h prior to operation of the QNPP Base,and is lower than the monitoring result of 101 nGy/h at the reference site in Hangzhou City.Within 50 km from the QNPP Base,the cumulative dose rate of the thermoluminescent dosimeter(TLD)is 90 nGy/h,which is lower than the back- ground level of 111 nGy/h.

Natural background radiation dose rate levels and incidences of reproductive abnormalities in high radiation area in Abeokuta, Southwestern Nigeria

A 10 y (1999-2008) birth records from two public and most accessible maternity hospitals locally in the city of Abeokuta, Nigeria were used to investigate the possible association of high outdoor gamma radiation exposure on reproductive abnormalities in the city. From the delivery record of 11,923 births in the period under study, a total number of 485 incidences of reproductive abnormalities were recoded. These incidences comprise 228 multiple births, 190 still births and, 67 premature births. Using the available terrestrial gamma radiation exposure data for the city and different reproductive abnormalities, regression assessment was carried out using the Pearson Product Moment (PPM) correlation statistics. The correlation showed that the incidences of reproductive abnormalities and the radiation dose levels were negatively correlated and correlation coefficient values were very low for each of the reproductive abnormalities considered. Factors such as socio-economic potentials of patients, dietary and other environmental factors may have substantial influence on the reproductive defects in the area other than radiation. However, the present study has added to the radiometric information needed in understanding the relationship between natural outdoor radiation exposure and occurrences of reproductive abnormalities in areas of high radiation in the country.

Evaluation of NORM and Dose Assessment in an Aluminium Industry in Nigeria

The activity concentrations of natural radionuclides 226Ra, 232Th and 40K in bauxite ore, alumina, dross tailing, aluminium scraps and soil samples collected from an aluminium industry in Nigeria were determined by gamma ray spectroscopy method. The mean values of 226Ra, 232Th and 40K content of the samples ranged from 16 ± 6 (alumina) to 31 ± 10 (scrap), 41 ± 0.12 (scrap) to 134 ± 21(bauxite) and 47 ± 14 (bauxite) to 354 ± 8 (scrap) Bq·kg-1, respectively. The mean activity concentrations of 226Ra and 40K in all the samples are lower than the world average for soil while 232Th is higher with the exception of alumina and scrap. As a measure of radiation hazard to the occupational workers and the members of the public, the radium equivalent activities and external gamma dose rates due to the radionuclides at 1 m above ground surface were calculated. The radium equivalent activities which varied between 88 ± 10 (alumina) and 222 ± 34 (bauxite) Bq·kg-1 are within the safety recommended limit of 370 Bq·kg-1. The mean annual effective doses calculated from the absorbed dose rates in air were between 54 ± 6 (alumina) and 134 ± 20 (bauxite) μSv·y-1, which is lower than the 1 mSv·y-1 recommended for the general public. The annual gonadal dose equivalent of all the samples with the exception of alumina was higher than the world average for soil.

Measurement of Gamma Emitting Radionuclides for Assessment, Environmental Hazards of Radiation in Rock and Soil Samples of Shabwah and Hadramout Regions, Yemen

The natural radioactivity of 226Ra, 232Th and 40K and the fallout of 137Cs in soils and granite rocks of two regions in Yemen (Shabwah and Hadramout) were measured by using gamma-ray spectrometry (HPGe) detector. The average values of the radionuclides 226Ra, 232Th and 40K in the soil samples are 14.34, 25.78 and 566.05 Bq/kg respectively. For rock samples, the average activity concentration for 226Ra, 232Th and 40K are 45, 106 and 1235 Bq/kg respectively. Low concentration values of 137Cs in soil and rock samples under investigation, are not radiologically important. The radium equivalent activity (Raeq) and external hazard index (Hex) of all samples are less than the limits of 370 Bq·kg-1 and unity, respectively. The average values of total absorbed dose rate due to three primordial radionuclides in soil and rock samples are 46.5 nGy/h and 138.36 nGy/h, respectively, where the absorbed average value of the granite is higher than the permitted limit. So, the local people must avoid using these granite samples as the interior decorative materials of dwelling without radioactivity control.

Radioactivity Measurements and Radiation Dose Assessments in Ground Water of Al-Baha Region, Saudi Arabia

Twenty samples of ground water were collected from different wells in Al-Baha region which is located at south-east of Saudi Arabia. Gamma spectrometer based on HPGe crystal was applied to determine activity concentrations in Bq·L-1 of the natural radio nuclides: Radium-226 and Thorium-232 series also Potassium-40. The measured results indicate that the average concentrations of 226Ra, 232Th and 40K in the studied samples were 0.85, 0.43, and 2.84 Bq·L-1, respectively. The average annual estimated effective dose was found to be 0.058 mSv/y which is lower than the annual limit of the dose allowed by WHO Thus, it has no harmful effects on health.

Assessment of Natural Radioactivity Levels and Radiation Dose Rate in Some Soil Samples from Historical Area, AL-RAKKAH, Saudi Arabia

This study aims to determine the activity concentrations of naturally occurring, technically-enhanced levels of radiation and the gamma absorbed dose rates in soil samples collected across the land scape of historical area which discovered in east of Saudi Arabia at 2009 G, Called AL- RAKKAH. By using an HPGe detector gamma-ray spectrometer, the activity concentrations of 226Ra 232Th and 40K were found in surface soil samples ranged from 17. 4 ± 1.2 Bq/kg to 28.3 ± 2.3 Bq/kg with an average value of 23 ± 1.6 Bq/kg, ranging from 1.1 ± 1.8 Bq/kg to 81.0 ± 1.7 Bq/kg with the average value 20 ± 1.4 Bq/kg and from 218 ± 11 Bq/kg to 255 ± 18 Bq/kg, with the mean value of 233 ± 12 Bq/kg respectively. The mean radium equivalent (Raeq) and outdoor radiation hazard index (Hex) for the area under study were determined as 69.52 Bq/kg and 0.16 respectively. The total absorbed dose rate due to three primordial radionuclides lies in the range of 17.74 - 72.24 nGy·h-1 with a mean of 32.69 nGy·h-1, which yields total annual effective dose of 0.37 mSv·y-1. The measured values are comparable with other global radioactivity measurements and are found to be safe for public and environment. The baseline data of this type will almost certainly be of importance in making estimations of population exposure.