Hariandra Muthu, Ramesh Kasi, R. Subramaniam, Shahid Bashir
The activity concentrations of naturally occurring radionuclides in vegetables and peat soil were investigated to determine the transfer factors of radionuclides from soil to vegetables obtained from farms in Klang Selangor. The results showed that the activity concentration ranges for 226Ra, 232Th, and 40K in the soil of the vegetable were 2.72-46.54 Bqkg-1, 9.01-54.84 Bqkg-1, and 19.22-477.76 Bqkg-1, respectively. The activity concentration ranges for 226Ra, 232Th, and 40K in various vegetable samples were 0.41-3.41 Bqkg-1, 0.02-3.56 Bqkg-1, and 16.22-317.49 Bqkg-1, respectively. The transfer factors from soil-to-plant for 226Ra, 232Th, and 40K were in the ranges of 0.01-0.67, 0.00-0.17, and 0.26-2.52, respectively. Radionuclide 40K has the highest transfer factor value in most vegetables, especially in the fruit type vegetable, and 232Th was found to have the lowest transfer factor value in all vegetables. The acquired results were compared to the levels that are globally recognized as acceptable. The transfer factor value of 40K was greater than 1 found in beans and fruit type vegetables, showing a similar value in other studies involving tropical plants, while transfer factor value for 226Ra was 15 times higher than the value reported in the data for tropical plants.
{"title":"Radioactivity concentration and transfer factors of natural radionuclides 226Ra, 232Th, and 40K from peat soil to vegetables in Selangor, Malaysia","authors":"Hariandra Muthu, Ramesh Kasi, R. Subramaniam, Shahid Bashir","doi":"10.2298/ntrp2201057m","DOIUrl":"https://doi.org/10.2298/ntrp2201057m","url":null,"abstract":"The activity concentrations of naturally occurring radionuclides in vegetables and peat soil were investigated to determine the transfer factors of radionuclides from soil to vegetables obtained from farms in Klang Selangor. The results showed that the activity concentration ranges for 226Ra, 232Th, and 40K in the soil of the vegetable were 2.72-46.54 Bqkg-1, 9.01-54.84 Bqkg-1, and 19.22-477.76 Bqkg-1, respectively. The activity concentration ranges for 226Ra, 232Th, and 40K in various vegetable samples were 0.41-3.41 Bqkg-1, 0.02-3.56 Bqkg-1, and 16.22-317.49 Bqkg-1, respectively. The transfer factors from soil-to-plant for 226Ra, 232Th, and 40K were in the ranges of 0.01-0.67, 0.00-0.17, and 0.26-2.52, respectively. Radionuclide 40K has the highest transfer factor value in most vegetables, especially in the fruit type vegetable, and 232Th was found to have the lowest transfer factor value in all vegetables. The acquired results were compared to the levels that are globally recognized as acceptable. The transfer factor value of 40K was greater than 1 found in beans and fruit type vegetables, showing a similar value in other studies involving tropical plants, while transfer factor value for 226Ra was 15 times higher than the value reported in the data for tropical plants.","PeriodicalId":49734,"journal":{"name":"Nuclear Technology & Radiation Protection","volume":"1 1","pages":""},"PeriodicalIF":1.2,"publicationDate":"2022-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"68548008","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
M. Razab, N. Nawi, R. Sunaiwi, A. Noor, M. Aziz, F. Hadzuan, Fathirah Ibrahim, A. Khaizul, N. Abdullah
Dealing with open sources of radioactive substances in nuclear medicine is a daily task since contamination due to radioactive spills may happen frequently. Proper and safe decontamination management is a vital procedure. However, regular purchase of decontamination agents incurs high costs and might be toxic due to their chemical properties. The purpose of this study is to compare graphene oxide, which is an environmentally friendly carbon-based material and marketable decontamination agent, in decontaminating radioactive spillage. Samples of pure 99mTc and 131I from the laboratory were spilled on a petri dish. The spill was immediately decontaminated with a marketable decontamination agent swab and varying concentrations of graphene oxide swab. The initial radioactivity of each swab containing 99mTc and 131I was measured using a dose calibrator. The absorbance spectra of each sample were analysed using an ultraviolet-visible spectrophotometer. The morphology image of graphene oxide was observed under field emission scanning electron microscope. For decontamination using a marketable decontamination agent, the radioactivity of 131I was slightly higher, whereas that of 99mTc was slightly lower than the high concentration of graphene oxide. The absorbance spectra of 99mTc and 131I that had been decontaminated using graphene oxide were observed at a range of 200 nm to 250 nm due ???* to the transition.
{"title":"Efficiency of marketable decontamination agent and graphene oxide on 99mTc and 131I spillages in nuclear medicine department","authors":"M. Razab, N. Nawi, R. Sunaiwi, A. Noor, M. Aziz, F. Hadzuan, Fathirah Ibrahim, A. Khaizul, N. Abdullah","doi":"10.2298/ntrp2202159r","DOIUrl":"https://doi.org/10.2298/ntrp2202159r","url":null,"abstract":"Dealing with open sources of radioactive substances in nuclear medicine is a daily task since contamination due to radioactive spills may happen frequently. Proper and safe decontamination management is a vital procedure. However, regular purchase of decontamination agents incurs high costs and might be toxic due to their chemical properties. The purpose of this study is to compare graphene oxide, which is an environmentally friendly carbon-based material and marketable decontamination agent, in decontaminating radioactive spillage. Samples of pure 99mTc and 131I from the laboratory were spilled on a petri dish. The spill was immediately decontaminated with a marketable decontamination agent swab and varying concentrations of graphene oxide swab. The initial radioactivity of each swab containing 99mTc and 131I was measured using a dose calibrator. The absorbance spectra of each sample were analysed using an ultraviolet-visible spectrophotometer. The morphology image of graphene oxide was observed under field emission scanning electron microscope. For decontamination using a marketable decontamination agent, the radioactivity of 131I was slightly higher, whereas that of 99mTc was slightly lower than the high concentration of graphene oxide. The absorbance spectra of 99mTc and 131I that had been decontaminated using graphene oxide were observed at a range of 200 nm to 250 nm due ???* to the transition.","PeriodicalId":49734,"journal":{"name":"Nuclear Technology & Radiation Protection","volume":"1 1","pages":""},"PeriodicalIF":1.2,"publicationDate":"2022-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"68548585","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
A. Kovacevic, N. Munic, Nenko Brkljač, K. Stanković
This paper presents an interlaboratory comparison of radiated emission measurements in the frequency range of 30-1000 MHz. A tubular dipole was specifically designed and employed as a reference emitting source. The most important for a tubular dipole is stability in the testing process. The stability is not the performance of the sample, but the emission signal of the sample is stable. In addition, two ways of determining the reference value of the electromagnetic field strength are considered. The first reference value is obtained by using robust analysis. It is a robust average value that is calculated by averaging the measurement results provided by the participating testing laboratories. The other reference value is obtained through the simulation-experimental results of the tubular dipole in the semi-anechoic chamber or full anechoic chamber, for horizontal and vertical polarizations and 3 m distance measurement, respectively. In addition, this value is assigned by the coordinator. Measurement results are compared using the robust z-scores and -scores, respectively.
{"title":"Interlaboratory comparison of radiated emission measurements using a tubular dipole","authors":"A. Kovacevic, N. Munic, Nenko Brkljač, K. Stanković","doi":"10.2298/ntrp2203207k","DOIUrl":"https://doi.org/10.2298/ntrp2203207k","url":null,"abstract":"This paper presents an interlaboratory comparison of radiated emission measurements in the frequency range of 30-1000 MHz. A tubular dipole was specifically designed and employed as a reference emitting source. The most important for a tubular dipole is stability in the testing process. The stability is not the performance of the sample, but the emission signal of the sample is stable. In addition, two ways of determining the reference value of the electromagnetic field strength are considered. The first reference value is obtained by using robust analysis. It is a robust average value that is calculated by averaging the measurement results provided by the participating testing laboratories. The other reference value is obtained through the simulation-experimental results of the tubular dipole in the semi-anechoic chamber or full anechoic chamber, for horizontal and vertical polarizations and 3 m distance measurement, respectively. In addition, this value is assigned by the coordinator. Measurement results are compared using the robust z-scores and -scores, respectively.","PeriodicalId":49734,"journal":{"name":"Nuclear Technology & Radiation Protection","volume":"1 1","pages":""},"PeriodicalIF":1.2,"publicationDate":"2022-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"68549302","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
V. Gostilo, A. Vlasenko, Vasily Litvinsky, Igors Krainukovs
The results of the development of modern precision monitors of alpha, beta and gamma ray radiation for setting up early warning systems for radioactive contamination in the atmosphere and rapid assessment of emerging threats, are presented. Proportional counters, scintillation SrI (Eu) crystals and semiconductor Si, CdZnTe, and HPGe detectors are used for 2 the development. The designed monitors provide information both on dose rate values in real time and on the activity of specific radionuclides. The software controls the measurement mode, as well as diagnoses the condition of the monitors themselves.
{"title":"Development of nuclear radiation monitors for radiation early warning systems","authors":"V. Gostilo, A. Vlasenko, Vasily Litvinsky, Igors Krainukovs","doi":"10.2298/ntrp2203193g","DOIUrl":"https://doi.org/10.2298/ntrp2203193g","url":null,"abstract":"The results of the development of modern precision monitors of alpha, beta and gamma ray radiation for setting up early warning systems for radioactive contamination in the atmosphere and rapid assessment of emerging threats, are presented. Proportional counters, scintillation SrI (Eu) crystals and semiconductor Si, CdZnTe, and HPGe detectors are used for 2 the development. The designed monitors provide information both on dose rate values in real time and on the activity of specific radionuclides. The software controls the measurement mode, as well as diagnoses the condition of the monitors themselves.","PeriodicalId":49734,"journal":{"name":"Nuclear Technology & Radiation Protection","volume":"1 1","pages":""},"PeriodicalIF":1.2,"publicationDate":"2022-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"68549019","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
This analysis of 238U, 226Ra and 210Pb transfer factors from the soil to the leaves of different native broadleaf trees at sites previously modified by uranium presence and at the site of background radioactivity levels, was conducted using data from a few available studies from the literature. The broadleaf tree species Quercus ilex, Quercus suber, Eucalyptus camaldulensis, Quercus pyrenaica, Quercus ilex rotundifolia, Populus sp. and Eucalyptus botryoides Sm. at the affected sites and Tilia spp. and Aesculus hippocastanum L. at the back ground site were in cluded in the study regardless of the deciduous or evergreen origins of the leaves. In the papers cited here, data about basic soil parameters: pH, total Ca [gkg-1], sand [%], and silt + clay [%] fractions were also available. All the collected data of activity concentration [Bqkg-1] dry weight in the soil (n=14) which was in the range: 22-6606 for 238U, 38-7700 for 226Ra, and 37-7500 for 210Pb, and the tree leaves in the range: 2.7-137.6 for 238U (n=10), 2.6-134.2 for 226Ra (n=14), and 27-77.2 for 210Pb (n=14), indicated that it was normally distributed after log-transformation. The present study was conducted under the hypothesis that biological differences between the examined broadleaf tree species have a lesser influence on the transfer factors of the investigated radionuclides from soil to tree leaves compared to the impact of the soil parameters and radionuclides activity concentrations in the soil. Consequently, it was examined whether 238U, 226Ra, and 210Pb soil-to-leaves transfer factor values for average broadleaf species could be predicted statistically in the first approximation based on their activity concentration in the soil and at least one basic soil parameter using multiple linear regression.
{"title":"Analysis of 238U, 226Ra, and 210Pb transfer factors from soil to the leaves of broadleaf tree species","authors":"I. Vukasinovic","doi":"10.2298/ntrp2203219v","DOIUrl":"https://doi.org/10.2298/ntrp2203219v","url":null,"abstract":"This analysis of 238U, 226Ra and 210Pb transfer factors from the soil to the leaves of different native broadleaf trees at sites previously modified by uranium presence and at the site of background radioactivity levels, was conducted using data from a few available studies from the literature. The broadleaf tree species Quercus ilex, Quercus suber, Eucalyptus camaldulensis, Quercus pyrenaica, Quercus ilex rotundifolia, Populus sp. and Eucalyptus botryoides Sm. at the affected sites and Tilia spp. and Aesculus hippocastanum L. at the back ground site were in cluded in the study regardless of the deciduous or evergreen origins of the leaves. In the papers cited here, data about basic soil parameters: pH, total Ca [gkg-1], sand [%], and silt + clay [%] fractions were also available. All the collected data of activity concentration [Bqkg-1] dry weight in the soil (n=14) which was in the range: 22-6606 for 238U, 38-7700 for 226Ra, and 37-7500 for 210Pb, and the tree leaves in the range: 2.7-137.6 for 238U (n=10), 2.6-134.2 for 226Ra (n=14), and 27-77.2 for 210Pb (n=14), indicated that it was normally distributed after log-transformation. The present study was conducted under the hypothesis that biological differences between the examined broadleaf tree species have a lesser influence on the transfer factors of the investigated radionuclides from soil to tree leaves compared to the impact of the soil parameters and radionuclides activity concentrations in the soil. Consequently, it was examined whether 238U, 226Ra, and 210Pb soil-to-leaves transfer factor values for average broadleaf species could be predicted statistically in the first approximation based on their activity concentration in the soil and at least one basic soil parameter using multiple linear regression.","PeriodicalId":49734,"journal":{"name":"Nuclear Technology & Radiation Protection","volume":"1 1","pages":""},"PeriodicalIF":1.2,"publicationDate":"2022-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"68549147","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The 1-D computer code MITH was used in this paper to perform sub-channel thermal-hydraulic analyses of a typical (Westinghouse model) pressurized water reactor. Two typical channels, hot and average, with the same flow rate and pressure drop, were tested under steady-state operating conditions. In this analysis, the channel with the highest temperature is designated as the hot channel. For the calculations, the channel model was divided into 20 parts. The thermal-hydraulic performance of the tested reactor was affected by power distribution, power level, and coolant mass-flow rate. Temperature distribution profiles of the fuel element and coolant are obtained for the average and hottest channels. A critical heat flux qncr analysis is also carried out and the heat fluxes in both channels were calculated. The W-3 correlation is employed to examine qncr in the hottest channel. Some data from the pressurized water reactor typical data sheet were used as input data, while others were used to validate the code. The code faithfully reproduced the Westinghouse model reactor results, including coolant, cladding, centerline, and surface fuel temperatures, quality and local heat flux qnloc, qncr and minimum departure from nucleate boiling ratio.
{"title":"Thermal-hydraulic analysis of light water reactors under different steady-state operating conditions, Part 2: Pressurized water reactor","authors":"E. Hutli, Ramadan Kridan","doi":"10.2298/ntrp2204276h","DOIUrl":"https://doi.org/10.2298/ntrp2204276h","url":null,"abstract":"The 1-D computer code MITH was used in this paper to perform sub-channel thermal-hydraulic analyses of a typical (Westinghouse model) pressurized water reactor. Two typical channels, hot and average, with the same flow rate and pressure drop, were tested under steady-state operating conditions. In this analysis, the channel with the highest temperature is designated as the hot channel. For the calculations, the channel model was divided into 20 parts. The thermal-hydraulic performance of the tested reactor was affected by power distribution, power level, and coolant mass-flow rate. Temperature distribution profiles of the fuel element and coolant are obtained for the average and hottest channels. A critical heat flux qncr analysis is also carried out and the heat fluxes in both channels were calculated. The W-3 correlation is employed to examine qncr in the hottest channel. Some data from the pressurized water reactor typical data sheet were used as input data, while others were used to validate the code. The code faithfully reproduced the Westinghouse model reactor results, including coolant, cladding, centerline, and surface fuel temperatures, quality and local heat flux qnloc, qncr and minimum departure from nucleate boiling ratio.","PeriodicalId":49734,"journal":{"name":"Nuclear Technology & Radiation Protection","volume":"1 1","pages":""},"PeriodicalIF":1.2,"publicationDate":"2022-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"68549479","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Busra Durmaz, Gizem Bakir, Bugra Arslan, H. Yapici
The study presents the analysis of the reusability of ThO2 and spent UO2 fuels enriched in two different ADS reactors fuelled with Minor Actinide. The spent UO2 fuels are taken out from pressurized water reactor and CANDU spent fuels. For this analysis, the CANDU-37 reactor having a total fission thermal power of 2156 MW is considered and 14 different cases of enriched fuels taken from the previous enrichment processes are analysed by burning in this reactor. The 3-D and time-dependent critical burn up calculations are carried out by using the MCNP 2.7 code. To determine the effective burn time of each case, these calculations are performed until the values of kinf decrease to about the criticality thresh old of 1.05 for all investigated cases. The percent ages of the 239Pu and 233U fissile isotopes appear to be below weapons-grade plutonium and uranium, respectively, in all enriched fuel cases. At the end of effective burn times, the burnup values can reach the values varying in the range of 26.770 and 33.540 GWd/MTU which are a mean of 3.5-4.5 times the burnup value of the CANDU-37 reactor fed with the NatUO2 fuel. The results of this study bring out that in terms of energy production, the CANDU-37 reactor fuelled with the ThO2 and spent UO2 fuels enriched in ADS designs demonstrates higher neutronic performance than the conventional CANDU-37 reactor.
{"title":"Analysis of reusability of ThO2 and spent UO2 fuels enriched with ads in a CANDU reactor","authors":"Busra Durmaz, Gizem Bakir, Bugra Arslan, H. Yapici","doi":"10.2298/ntrp2204289d","DOIUrl":"https://doi.org/10.2298/ntrp2204289d","url":null,"abstract":"The study presents the analysis of the reusability of ThO2 and spent UO2 fuels enriched in two different ADS reactors fuelled with Minor Actinide. The spent UO2 fuels are taken out from pressurized water reactor and CANDU spent fuels. For this analysis, the CANDU-37 reactor having a total fission thermal power of 2156 MW is considered and 14 different cases of enriched fuels taken from the previous enrichment processes are analysed by burning in this reactor. The 3-D and time-dependent critical burn up calculations are carried out by using the MCNP 2.7 code. To determine the effective burn time of each case, these calculations are performed until the values of kinf decrease to about the criticality thresh old of 1.05 for all investigated cases. The percent ages of the 239Pu and 233U fissile isotopes appear to be below weapons-grade plutonium and uranium, respectively, in all enriched fuel cases. At the end of effective burn times, the burnup values can reach the values varying in the range of 26.770 and 33.540 GWd/MTU which are a mean of 3.5-4.5 times the burnup value of the CANDU-37 reactor fed with the NatUO2 fuel. The results of this study bring out that in terms of energy production, the CANDU-37 reactor fuelled with the ThO2 and spent UO2 fuels enriched in ADS designs demonstrates higher neutronic performance than the conventional CANDU-37 reactor.","PeriodicalId":49734,"journal":{"name":"Nuclear Technology & Radiation Protection","volume":"1 1","pages":""},"PeriodicalIF":1.2,"publicationDate":"2022-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"68550633","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
This work presents a development of a new radiocarbon dating system based on CO2 and liquid scintillation counting. The system is flexible enough to be applied in different fields such as radiocarbon dating (up to about 32 000 years span), environmental radioactive monitoring, and determining the fraction of biological carbon in environmentally-ecologically safe biodegradable and bio-synthetic products, and help accreditation to customaries. The implemented system is developed following the standard method (ASTM D6866-06). It has been calibrated tested and verified for CO2 production, absorption, and counting (using appropriate reference materials for the present-day, fossil, and in between ages). These results are optimistic that the developed system can contribute to the accreditation of ongoing and upcoming environment- friendly productions.
{"title":"Development of direct CO2 absorption system for radiocarbon dating and bio-based carbon determination in biogenic-synthetic products","authors":"M. Sadek, Raafat Rayan, Waleed F. Khalil","doi":"10.2298/ntrp2204315s","DOIUrl":"https://doi.org/10.2298/ntrp2204315s","url":null,"abstract":"This work presents a development of a new radiocarbon dating system based on CO2 and liquid scintillation counting. The system is flexible enough to be applied in different fields such as radiocarbon dating (up to about 32 000 years span), environmental radioactive monitoring, and determining the fraction of biological carbon in environmentally-ecologically safe biodegradable and bio-synthetic products, and help accreditation to customaries. The implemented system is developed following the standard method (ASTM D6866-06). It has been calibrated tested and verified for CO2 production, absorption, and counting (using appropriate reference materials for the present-day, fossil, and in between ages). These results are optimistic that the developed system can contribute to the accreditation of ongoing and upcoming environment- friendly productions.","PeriodicalId":49734,"journal":{"name":"Nuclear Technology & Radiation Protection","volume":"1 1","pages":""},"PeriodicalIF":1.2,"publicationDate":"2022-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"68550890","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
A. Osman, B. El, Samad El, Z. Alsayed, R. Awad, M. Badawi
Lead oxide (PbO) bulk and nanoparticles of two different sizes (A = 78 nm and B = 54 nm) are incorporated separately into the polystyrene matrix at various concentrations (0, 10, 15, 25, and 35 %) using roll mill mixing and compressing molding techniques. The X-ray narrow-spectrum series (N-series / ISO 4037-1) is then used to investigate the radiation attenuation capability of the novel polymer composite PS/PbO, as well as the effect of varying PbO particle sizes on shielding performance. The filler dispersion and chemical elemental analysis of the synthesized composite are investigated using scanning electron microscopy and energy-dispersive X-ray spectroscopy. To determine the mass attenuation coefficients ?m, samples with various thicknesses of the synthesized composite are examined using a range of X-ray energies, and the experimental data are compared to theoretical values from NIST databases (XCOM and FFAST). The results indicate that either increasing the filler weight percentage or, decreasing the filler particle size, enhanced the attenuation parameters throughout all energies. The composite containing the smallest nanosize of PbO exhibited the maximum radiation shielding efficacy among all combinations and therefore, might be used to develop low-cost and lightweight X-ray shields.
采用辊磨混合和压缩成型技术，将两种不同尺寸(A = 78 nm和B = 54 nm)的氧化铅(PbO)体和纳米颗粒分别以不同浓度(0%、10%、15%、25%和35%)掺入聚苯乙烯基体中。利用x射线窄谱系列(n -系列/ ISO 4037-1)研究了新型聚合物复合材料PS/PbO的辐射衰减能力，以及不同PbO粒径对屏蔽性能的影响。利用扫描电镜和能量色散x射线能谱对合成的复合材料进行了填料弥散和化学元素分析。为了确定质量衰减系数?m，利用x射线能量范围检测了不同厚度的合成复合材料样品，并将实验数据与NIST数据库(XCOM和FFAST)的理论值进行了比较。结果表明，无论是增加填料的重量百分比，还是减小填料的粒径，都能增强各能量的衰减参数。含PbO纳米尺寸最小的复合材料在所有组合中表现出最大的辐射屏蔽效果，因此可用于开发低成本、轻量化的x射线屏蔽材料。
{"title":"Effect of PbO incorporation with different particle size on X-ray attenuation of polystyrene","authors":"A. Osman, B. El, Samad El, Z. Alsayed, R. Awad, M. Badawi","doi":"10.2298/ntrp2201018o","DOIUrl":"https://doi.org/10.2298/ntrp2201018o","url":null,"abstract":"Lead oxide (PbO) bulk and nanoparticles of two different sizes (A = 78 nm and B = 54 nm) are incorporated separately into the polystyrene matrix at various concentrations (0, 10, 15, 25, and 35 %) using roll mill mixing and compressing molding techniques. The X-ray narrow-spectrum series (N-series / ISO 4037-1) is then used to investigate the radiation attenuation capability of the novel polymer composite PS/PbO, as well as the effect of varying PbO particle sizes on shielding performance. The filler dispersion and chemical elemental analysis of the synthesized composite are investigated using scanning electron microscopy and energy-dispersive X-ray spectroscopy. To determine the mass attenuation coefficients ?m, samples with various thicknesses of the synthesized composite are examined using a range of X-ray energies, and the experimental data are compared to theoretical values from NIST databases (XCOM and FFAST). The results indicate that either increasing the filler weight percentage or, decreasing the filler particle size, enhanced the attenuation parameters throughout all energies. The composite containing the smallest nanosize of PbO exhibited the maximum radiation shielding efficacy among all combinations and therefore, might be used to develop low-cost and lightweight X-ray shields.","PeriodicalId":49734,"journal":{"name":"Nuclear Technology & Radiation Protection","volume":"1 1","pages":""},"PeriodicalIF":1.2,"publicationDate":"2022-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"68547460","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The deuterium-tritium neutron generator should be fully shielded for the safety of the operators participating in the experiments since the D-T neutron generator is commonly used in activation experiments. In this study, MCNP5 code was used to simulate the shielding effect of the neutron thermalization device previously designed by our group with Pb and boron-containing polyethylene as the shielding material. The neutron dose rate outside of the previous thermalization device can not meet the requirement, so a concrete wall is needed between the device and the operators. Two models are designed with concrete walls. One model is that the device and the experimental operators are not in the same room, another one is that the device and the experimental operators are in the same room, and there is an L-shaped concrete wall between them. In both models, the dose rate to the operators was less than 5 ?Svh-1.
{"title":"Shielding design in neutron activation experiment system based on D-T neutron tube","authors":"Jingfei Cai, Shiwei Jing, Dedong He, Yangwenting Ou, Xinyi Ling, Bingbing Li","doi":"10.2298/ntrp2201042c","DOIUrl":"https://doi.org/10.2298/ntrp2201042c","url":null,"abstract":"The deuterium-tritium neutron generator should be fully shielded for the safety of the operators participating in the experiments since the D-T neutron generator is commonly used in activation experiments. In this study, MCNP5 code was used to simulate the shielding effect of the neutron thermalization device previously designed by our group with Pb and boron-containing polyethylene as the shielding material. The neutron dose rate outside of the previous thermalization device can not meet the requirement, so a concrete wall is needed between the device and the operators. Two models are designed with concrete walls. One model is that the device and the experimental operators are not in the same room, another one is that the device and the experimental operators are in the same room, and there is an L-shaped concrete wall between them. In both models, the dose rate to the operators was less than 5 ?Svh-1.","PeriodicalId":49734,"journal":{"name":"Nuclear Technology & Radiation Protection","volume":"1 1","pages":""},"PeriodicalIF":1.2,"publicationDate":"2022-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"68547676","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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