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}
As a historic challenge for humans, Martian colonization has been initiated by nuclear energy. A moving nuclear power plant could be imaginable known as a nuclear reactor rover. The design of the nuclear reactor rover has been performed where the important matter is how to make the caterpillar move the reactor and its facilities. Hence the slider length and contact point are proposed. The normalized heat transfer is analyzed by slide length and contact point where they are normalized as 1.0 and 10.0, respectively. Although the slider length of the caterpillar is proportional to heat transfer, the contact point shows the adverse values. Longer slider length and less contact point could be the optimized heat production system by the caterpillar which is the additional heat source except the other nuclear reactor. Any other planet could be considered as a potential human colony using the nuclear terraforming technology.
{"title":"Analysis of terraforming on mars using nuclear power for the beginning of space colonization","authors":"T. Woo, C. Baek, K. Jang","doi":"10.2298/ntrp2203253w","DOIUrl":"https://doi.org/10.2298/ntrp2203253w","url":null,"abstract":"As a historic challenge for humans, Martian colonization has been initiated by nuclear energy. A moving nuclear power plant could be imaginable known as a nuclear reactor rover. The design of the nuclear reactor rover has been performed where the important matter is how to make the caterpillar move the reactor and its facilities. Hence the slider length and contact point are proposed. The normalized heat transfer is analyzed by slide length and contact point where they are normalized as 1.0 and 10.0, respectively. Although the slider length of the caterpillar is proportional to heat transfer, the contact point shows the adverse values. Longer slider length and less contact point could be the optimized heat production system by the caterpillar which is the additional heat source except the other nuclear reactor. Any other planet could be considered as a potential human colony using the nuclear terraforming technology.","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":"68549790","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 steady-state thermal-hydraulic analysis of the core of the Boiling Water Reactor (BWR/6) at nominal operating conditions is presented in this paper. The BWR/6 is produced by General Electric USA. The analysis' goal is to keep the thermal safety margin under control and the core integrity intact under steady-state operating conditions. The effects of operating conditions such as power distribution, power level, and coolant mass flow rate on the pro- posed core's performance are investigated. For this purpose, the one-dimensional computer code MITH was used. The code's reliability was tested using the General Electric benchmark 3579 MW reactor. Two-channel models were tested (the average and the hot channel). Ther- mal-hydraulic parameters such as fuel-centerline, fuel-surface, outer clad surface and coolant temperature, critical and actual local heat flux, critical and minimum critical heat flux ratio and pressure drop are evaluated along the tested channels. Temperatures, as well as actual and critical heat flux distribution profiles, were obtained. The tested operating conditions had a significant influence on these parameters, and also on the thermal-hydraulic performance. The obtained results are in good agreement with the data from the tested core. The obtained results are well within the safety margins. The good agreement between tested reactor data and MITH code calculation concerning the reactor demonstrates the reliability of the analysis methodology from a thermal-hydraulic perspective.
{"title":"Thermal-hydraulic analysis of light water reactors under different steady-state operating conditions, Part 1: Boiling water reactor","authors":"E. Hutli, Ramadan Kridan","doi":"10.2298/ntrp2204259h","DOIUrl":"https://doi.org/10.2298/ntrp2204259h","url":null,"abstract":"The steady-state thermal-hydraulic analysis of the core of the Boiling Water Reactor (BWR/6) at nominal operating conditions is presented in this paper. The BWR/6 is produced by General Electric USA. The analysis' goal is to keep the thermal safety margin under control and the core integrity intact under steady-state operating conditions. The effects of operating conditions such as power distribution, power level, and coolant mass flow rate on the pro- posed core's performance are investigated. For this purpose, the one-dimensional computer code MITH was used. The code's reliability was tested using the General Electric benchmark 3579 MW reactor. Two-channel models were tested (the average and the hot channel). Ther- mal-hydraulic parameters such as fuel-centerline, fuel-surface, outer clad surface and coolant temperature, critical and actual local heat flux, critical and minimum critical heat flux ratio and pressure drop are evaluated along the tested channels. Temperatures, as well as actual and critical heat flux distribution profiles, were obtained. The tested operating conditions had a significant influence on these parameters, and also on the thermal-hydraulic performance. The obtained results are in good agreement with the data from the tested core. The obtained results are well within the safety margins. The good agreement between tested reactor data and MITH code calculation concerning the reactor demonstrates the reliability of the analysis methodology from a thermal-hydraulic perspective.","PeriodicalId":49734,"journal":{"name":"Nuclear Technology & Radiation Protection","volume":"290 1","pages":""},"PeriodicalIF":1.2,"publicationDate":"2022-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"68549865","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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