Aaron M. Chalifoux, Logan Gibb, Kimberly N. Wurth, Travis Tenner, Tolga Tasdizen, Luther W. McDonald
Morphological analysis of uranium materials has proven to be a key signature for nuclear forensic purposes. This study examines the morphological changes to magnesium diuranate (MDU) and sodium diuranate (SDU) during reduction in a 10 % hydrogen atmosphere with and without steam present. Impurity concentrations of the materials were also examined pre and post reduction using energy dispersive X-ray spectroscopy combined with scanning electron microscopy (SEM-EDX). The structures of the MDU, SDU, and UOx samples were analyzed using powder X-ray diffraction (p-XRD). Using this method, UOx from MDU was found to be a mixture of UO2, U4O9, and MgU2O6 while UOx from SDU were combinations of UO2, U4O9, U3O8, and UO3. By SEM, the MDU and UOx from MDU had identical morphologies comprised of large agglomerates of rounded particles in an irregular pattern. SEM-EDX revealed pockets of high U and high Mg content distributed throughout the materials. The SDU and UOx from SDU had slightly different morphologies. The SDU consisted of massive agglomerates of platy sheets with rough surfaces. The UOx from SDU was comprised of massive agglomerates of acicular and sub-rounded particles that appeared slightly sintered. Backscatter images of SDU and related UOx materials showed sub-rounded dark spots indicating areas of high Na content, especially in UOx materials created in the presence of steam. SEM-EDX confirmed the presence of high sodium concentration spots in the SDU and UOx from SDU. Elemental compositions were found to not change between pre and post reduction of MDU and SDU indicating that reduction with or without steam does not affect Mg or Na concentrations. The identification of Mg and Na impurities using SEM analysis presents a readily accessible tool in nuclear material analysis with high Mg and Na impurities likely indicating processing via MDU or SDU, respectively. Machine learning using convolutional neural networks (CNNs) found that the MDU and SDU had unique morphologies compared to previous publications and that there are distinguishing features between materials created with and without steam.
铀材料的形态分析已被证明是核鉴识目的的关键特征。本研究探讨了二呋喃酸镁（MDU）和二呋喃酸钠（SDU）在 10% 氢气环境中进行还原（有蒸汽存在和无蒸汽存在）时的形态变化。此外，还使用能量色散 X 射线光谱和扫描电子显微镜（SEM-EDX）对还原前后材料的杂质浓度进行了检测。使用粉末 X 射线衍射 (p-XRD) 分析了 MDU、SDU 和 UO x 样品的结构。通过这种方法，发现 MDU 中的氧化亚氮是二氧化硫、氧化亚氮和氧化镁的混合物，而 SDU 中的氧化亚氮是二氧化硫、氧化亚氮、氧化亚氮和氧化亚氮的混合物。通过扫描电子显微镜，MDU 和来自 MDU 的氧化铀具有相同的形态，都是由不规则的圆形颗粒组成的大团块。SEM-EDX 显示，在整个材料中分布着高铀和高镁含量的区域。来自 SDU 的 SDU 和 UO x 的形态略有不同。SDU由表面粗糙的板状片材组成。而从飞毛腿中提取的氧化亚铀则是由针状和亚圆形颗粒组成的大块团块，看起来略微烧结。SDU 和相关 UO x 材料的背向散射图像显示出亚圆形黑点，表明 Na 含量较高的区域，尤其是在蒸汽中生成的 UO x 材料中。SEM-EDX 证实了在 SDU 和来自 SDU 的 UO x 中存在高钠浓度斑点。元素组成在 MDU 和 SDU 还原前和还原后没有发生变化，这表明使用或不使用蒸汽进行还原不会影响镁或钠的浓度。利用扫描电子显微镜分析鉴定镁和钠杂质为核材料分析提供了一个易于使用的工具，高镁和高钠杂质可能分别表明是通过 MDU 或 SDU 加工的。利用卷积神经网络（CNN）进行的机器学习发现，与以前的出版物相比，MDU 和 SDU 具有独特的形态，而且在使用蒸汽和不使用蒸汽的情况下生成的材料之间存在区别特征。
{"title":"Morphology of uranium oxides reduced from magnesium and sodium diuranate","authors":"Aaron M. Chalifoux, Logan Gibb, Kimberly N. Wurth, Travis Tenner, Tolga Tasdizen, Luther W. McDonald","doi":"10.1515/ract-2023-0221","DOIUrl":"https://doi.org/10.1515/ract-2023-0221","url":null,"abstract":"Morphological analysis of uranium materials has proven to be a key signature for nuclear forensic purposes. This study examines the morphological changes to magnesium diuranate (MDU) and sodium diuranate (SDU) during reduction in a 10 % hydrogen atmosphere with and without steam present. Impurity concentrations of the materials were also examined pre and post reduction using energy dispersive X-ray spectroscopy combined with scanning electron microscopy (SEM-EDX). The structures of the MDU, SDU, and UO<jats:sub> <jats:italic>x</jats:italic> </jats:sub> samples were analyzed using powder X-ray diffraction (p-XRD). Using this method, UO<jats:sub> <jats:italic>x</jats:italic> </jats:sub> from MDU was found to be a mixture of UO<jats:sub>2</jats:sub>, U<jats:sub>4</jats:sub>O<jats:sub>9</jats:sub>, and MgU<jats:sub>2</jats:sub>O<jats:sub>6</jats:sub> while UO<jats:sub> <jats:italic>x</jats:italic> </jats:sub> from SDU were combinations of UO<jats:sub>2</jats:sub>, U<jats:sub>4</jats:sub>O<jats:sub>9</jats:sub>, U<jats:sub>3</jats:sub>O<jats:sub>8</jats:sub>, and UO<jats:sub>3</jats:sub>. By SEM, the MDU and UO<jats:sub> <jats:italic>x</jats:italic> </jats:sub> from MDU had identical morphologies comprised of large agglomerates of rounded particles in an irregular pattern. SEM-EDX revealed pockets of high U and high Mg content distributed throughout the materials. The SDU and UO<jats:sub> <jats:italic>x</jats:italic> </jats:sub> from SDU had slightly different morphologies. The SDU consisted of massive agglomerates of platy sheets with rough surfaces. The UO<jats:sub> <jats:italic>x</jats:italic> </jats:sub> from SDU was comprised of massive agglomerates of acicular and sub-rounded particles that appeared slightly sintered. Backscatter images of SDU and related UO<jats:sub> <jats:italic>x</jats:italic> </jats:sub> materials showed sub-rounded dark spots indicating areas of high Na content, especially in UO<jats:sub> <jats:italic>x</jats:italic> </jats:sub> materials created in the presence of steam. SEM-EDX confirmed the presence of high sodium concentration spots in the SDU and UO<jats:sub> <jats:italic>x</jats:italic> </jats:sub> from SDU. Elemental compositions were found to not change between pre and post reduction of MDU and SDU indicating that reduction with or without steam does not affect Mg or Na concentrations. The identification of Mg and Na impurities using SEM analysis presents a readily accessible tool in nuclear material analysis with high Mg and Na impurities likely indicating processing via MDU or SDU, respectively. Machine learning using convolutional neural networks (CNNs) found that the MDU and SDU had unique morphologies compared to previous publications and that there are distinguishing features between materials created with and without steam.","PeriodicalId":21167,"journal":{"name":"Radiochimica Acta","volume":null,"pages":null},"PeriodicalIF":1.8,"publicationDate":"2023-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139053549","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"化学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The surface soil samples were collected from Northwest Turkey. The activity concentrations of 226Ra, 232Th, and 40K were measured using an HPGe gamma-spectroscopy system. The activity concentrations of 226Ra, 232Th, and 40K in the soils were found to be in the range of 11.78 ± 1.12–43.89 ± 14.94, 3.19 ± 2.01–88.22 ± 0.92, 362.81 ± 5.94–829.27 ± 12.38 Bq kg−1 d.w., respectively. The Surfer program was used to obtain 3-dimensional maps of the specific activities. Radium Equivalent Activity (Raeq), Absorbed Gamma Dose Rate (D), Annual Effective Dose Equivalent (AEDE), The Excess Life Time Cancer Risk (ELCR), External (Hex) and Internal (Hin) Hazard Indexes, Annual Gonadal Dose Equivalent (AGDE), and Activity Utilization Index (AUI) were calculated and compared with the recommended values. Pearson’s correlation analysis (PCA) and factor analysis (FA) were utilized to analyze the data and indicate between the radiological parameters. The analysis showed that the total radiation was mainly caused by 226Ra and 232Th.
{"title":"Radioactivity of 226Ra, 232Th and 40K in soil in Northwest part of Turkey: assessment of radiological impacts","authors":"Selin Özden","doi":"10.1515/ract-2023-0219","DOIUrl":"https://doi.org/10.1515/ract-2023-0219","url":null,"abstract":"The surface soil samples were collected from Northwest Turkey. The activity concentrations of <jats:sup>226</jats:sup>Ra, <jats:sup>232</jats:sup>Th, and <jats:sup>40</jats:sup>K were measured using an HPGe gamma-spectroscopy system. The activity concentrations of <jats:sup>226</jats:sup>Ra, <jats:sup>232</jats:sup>Th, and <jats:sup>40</jats:sup>K in the soils were found to be in the range of 11.78 ± 1.12–43.89 ± 14.94, 3.19 ± 2.01–88.22 ± 0.92, 362.81 ± 5.94–829.27 ± 12.38 Bq kg<jats:sup>−1</jats:sup> d.w., respectively. The Surfer program was used to obtain 3-dimensional maps of the specific activities. Radium Equivalent Activity (Ra<jats:sub>eq</jats:sub>), Absorbed Gamma Dose Rate (<jats:italic>D</jats:italic>), Annual Effective Dose Equivalent (AEDE), The Excess Life Time Cancer Risk (ELCR), External (<jats:italic>H</jats:italic> <jats:sub>ex</jats:sub>) and Internal (<jats:italic>H</jats:italic> <jats:sub>in</jats:sub>) Hazard Indexes, Annual Gonadal Dose Equivalent (AGDE), and Activity Utilization Index (AUI) were calculated and compared with the recommended values. Pearson’s correlation analysis (PCA) and factor analysis (FA) were utilized to analyze the data and indicate between the radiological parameters. The analysis showed that the total radiation was mainly caused by <jats:sup>226</jats:sup>Ra and <jats:sup>232</jats:sup>Th.","PeriodicalId":21167,"journal":{"name":"Radiochimica Acta","volume":null,"pages":null},"PeriodicalIF":1.8,"publicationDate":"2023-12-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"138630784","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"化学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Yarasi Balaji Rao, Vinod K. Ray, Putta V. Nagendra Kumar, Dinesh Srivastava
Uranium concentration and uranium isotopic content are two important and critical parameters for any nuclear fuel fabrication facilities. In the present study emphasis has been given on the usage of high resolution gamma ray spectrometric (HR-GRS) technique with HPGe detector for the determination of uranium and 235U content in uranium process stream samples. The work has been carried out with an aim to give quick analytical feedback to production facility and also to minimize the generation of analytical waste. These are important requirements for any industrial lab with high analytical load attached to production facility. In this paper a simple and a non-destructive testing (NDT) method has been described for quantification of uranium and 235U content in samples received from UO2 fuel production facilities using HR-GRS technique with high purity germanium (HPGe) detector. A 185.7 keV line of 235U has been used for quantification of uranium in process solutions containing 1 g/L to 600 g/L of uranium covering both aqueous and organic process stream solutions. The results have been compared with that of Davies and Gray method. The limitations associated with gamma ray emitted from daughter products of 238U and self-induced or self-excited X-ray fluorescence lines of uranium have also been studied. Multi Group Analysis of Uranium (MGAU) software code has been used for measurement of 235U content in variety of samples. The results obtained are compared with that of results by thermal ionization mass spectrometer (TIMS).
{"title":"HR GRS-HPGe as NDT method for quantification of uranium and U235 content in process stream samples from UO2 fuel production facilities","authors":"Yarasi Balaji Rao, Vinod K. Ray, Putta V. Nagendra Kumar, Dinesh Srivastava","doi":"10.1515/ract-2023-0186","DOIUrl":"https://doi.org/10.1515/ract-2023-0186","url":null,"abstract":"Uranium concentration and uranium isotopic content are two important and critical parameters for any nuclear fuel fabrication facilities. In the present study emphasis has been given on the usage of high resolution gamma ray spectrometric (HR-GRS) technique with HPGe detector for the determination of uranium and <jats:sup>235</jats:sup>U content in uranium process stream samples. The work has been carried out with an aim to give quick analytical feedback to production facility and also to minimize the generation of analytical waste. These are important requirements for any industrial lab with high analytical load attached to production facility. In this paper a simple and a non-destructive testing (NDT) method has been described for quantification of uranium and <jats:sup>235</jats:sup>U content in samples received from UO<jats:sub>2</jats:sub> fuel production facilities using HR-GRS technique with high purity germanium (HPGe) detector. A 185.7 keV line of <jats:sup>235</jats:sup>U has been used for quantification of uranium in process solutions containing 1 g/L to 600 g/L of uranium covering both aqueous and organic process stream solutions. The results have been compared with that of Davies and Gray method. The limitations associated with gamma ray emitted from daughter products of <jats:sup>238</jats:sup>U and self-induced or self-excited X-ray fluorescence lines of uranium have also been studied. Multi Group Analysis of Uranium (MGAU) software code has been used for measurement of <jats:sup>235</jats:sup>U content in variety of samples. The results obtained are compared with that of results by thermal ionization mass spectrometer (TIMS).","PeriodicalId":21167,"journal":{"name":"Radiochimica Acta","volume":null,"pages":null},"PeriodicalIF":1.8,"publicationDate":"2023-12-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"138576755","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"化学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
In this contribution, the distribution of naturally occurring radionuclides in the area around Main Karakoram Thrust (MKT) in Karakoram Range, North Pakistan is documented. Three natural radionuclides (226Ra, 232Th, and 40K) and one anthropogenic radionuclide (137Cs) were studied for their specific activities in 30 samples. The measurements were made by high resolution gamma-ray spectrometry. The sampling area is located in Gilgit Baltistan province of Pakistan at an altitude of 1838 m/6030 ft above sea level. MKT separates the Karakoram plate from the Kohistan-Ladakh Terranes and Indian Plate to the south. The specific activity varied as 4.5–56.5 Bq kg−1, 18.2–61.4 Bq kg−1, 1.4–19.6 Bq kg−1 and 51–1640 Bq kg−1 for 226Ra, 232Th, 137Cs and 40K, respectively. The average radium equivalent activity was 127.8 ± 45.9 Bq kg−1. The external hazard index was <1 for all samples and representative level index was <1 for majority of the samples. The average air absorbed dose rate was 60.9 ± 23.2 nGy h−1 corresponding to the outdoor effective dose rate of 73.7 ± 28.0 μSv y−1. These values were slightly higher than the world average values for air absorbed dose rate (51 nGy h−1) and outdoor annual effective dose rate (70 μSv y−1). The data revealed significant positive correlation between 226Ra and 40K. Principal component analysis revealed distribution patterns within the samples and identified three distinct groups. Data was also evaluated for the concentrations of uranium, thorium and potassium and their ratios.
{"title":"Gamma-radiation levels along the main Karakorum thrust area of Northern Pakistan","authors":"Mohammad Wasim, Arfan Tariq, Manzoor Ali","doi":"10.1515/ract-2023-0229","DOIUrl":"https://doi.org/10.1515/ract-2023-0229","url":null,"abstract":"In this contribution, the distribution of naturally occurring radionuclides in the area around Main Karakoram Thrust (MKT) in Karakoram Range, North Pakistan is documented. Three natural radionuclides (<jats:sup>226</jats:sup>Ra, <jats:sup>232</jats:sup>Th, and <jats:sup>40</jats:sup>K) and one anthropogenic radionuclide (<jats:sup>137</jats:sup>Cs) were studied for their specific activities in 30 samples. The measurements were made by high resolution gamma-ray spectrometry. The sampling area is located in Gilgit Baltistan province of Pakistan at an altitude of 1838 m/6030 ft above sea level. MKT separates the Karakoram plate from the Kohistan-Ladakh Terranes and Indian Plate to the south. The specific activity varied as 4.5–56.5 Bq kg<jats:sup>−1</jats:sup>, 18.2–61.4 Bq kg<jats:sup>−1</jats:sup>, 1.4–19.6 Bq kg<jats:sup>−1</jats:sup> and 51–1640 Bq kg<jats:sup>−1</jats:sup> for <jats:sup>226</jats:sup>Ra, <jats:sup>232</jats:sup>Th, <jats:sup>137</jats:sup>Cs and <jats:sup>40</jats:sup>K, respectively. The average radium equivalent activity was 127.8 ± 45.9 Bq kg<jats:sup>−1</jats:sup>. The external hazard index was <1 for all samples and representative level index was <1 for majority of the samples. The average air absorbed dose rate was 60.9 ± 23.2 nGy h<jats:sup>−1</jats:sup> corresponding to the outdoor effective dose rate of 73.7 ± 28.0 μSv y<jats:sup>−1</jats:sup>. These values were slightly higher than the world average values for air absorbed dose rate (51 nGy h<jats:sup>−1</jats:sup>) and outdoor annual effective dose rate (70 μSv y<jats:sup>−1</jats:sup>). The data revealed significant positive correlation between <jats:sup>226</jats:sup>Ra and <jats:sup>40</jats:sup>K. Principal component analysis revealed distribution patterns within the samples and identified three distinct groups. Data was also evaluated for the concentrations of uranium, thorium and potassium and their ratios.","PeriodicalId":21167,"journal":{"name":"Radiochimica Acta","volume":null,"pages":null},"PeriodicalIF":1.8,"publicationDate":"2023-12-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"138553158","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"化学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The aim of this study is to create a radiological map of Kırşehir province in Turkey by analyzing the radioactivity levels of soil samples and environmental radiation measurements. 226Ra, 232Th, 40K and 137Cs radionuclides were analysed in soil samples collected from 47 locations using gamma spectrometry with an HPGe detector. According to the results obtained, gamma radioactivity concentration ranges and mean values for 226Ra, 232Th, 40K and 137Cs were determined as 133 ± 15–1515 ± 128 Bq/kg (599.40 Bq/kg), 0.3 ± 0.09–21.1 ± 1.7 Bq/kg (4.61 Bq/kg), 8.7 ± 0.9–128.5 ± 8.5 Bq/kg (29.66 Bq/kg) and 11.6 ± 8.6–273.8 ± 19.9 Bq/kg (48.80 Bq/kg) respectively. Average radioactivity levels for 40K and 232Th are above the world average. The mean radium equivalent activity (Raeq), gamma dose rate (D), annual effective dose equivalent (AEDE), external hazard index (Hex), and the excess lifetime cancer risk (ELCR) were calculated as 145.45 Bq/kg, 122.13 nGy/h, 0.15 mSv/h, 0.39 and 0.29 × 10−3 respectively.
{"title":"Determination of natural and artificial radioactivity levels and radiation hazard indices for soil samples in Kırşehir","authors":"Sümeyra Yamçıçıer, Doğan Yaşar","doi":"10.1515/ract-2023-0215","DOIUrl":"https://doi.org/10.1515/ract-2023-0215","url":null,"abstract":"The aim of this study is to create a radiological map of Kırşehir province in Turkey by analyzing the radioactivity levels of soil samples and environmental radiation measurements. <jats:sup>226</jats:sup>Ra, <jats:sup>232</jats:sup>Th, <jats:sup>40</jats:sup>K and <jats:sup>137</jats:sup>Cs radionuclides were analysed in soil samples collected from 47 locations using gamma spectrometry with an HPGe detector. According to the results obtained, gamma radioactivity concentration ranges and mean values for <jats:sup>226</jats:sup>Ra, <jats:sup>232</jats:sup>Th, <jats:sup>40</jats:sup>K and <jats:sup>137</jats:sup>Cs were determined as 133 ± 15–1515 ± 128 Bq/kg (599.40 Bq/kg), 0.3 ± 0.09–21.1 ± 1.7 Bq/kg (4.61 Bq/kg), 8.7 ± 0.9–128.5 ± 8.5 Bq/kg (29.66 Bq/kg) and 11.6 ± 8.6–273.8 ± 19.9 Bq/kg (48.80 Bq/kg) respectively. Average radioactivity levels for <jats:sup>40</jats:sup>K and <jats:sup>232</jats:sup>Th are above the world average. The mean radium equivalent activity (Ra<jats:sub>eq</jats:sub>), gamma dose rate (<jats:italic>D</jats:italic>), annual effective dose equivalent (AEDE), external hazard index (<jats:italic>H</jats:italic> <jats:sub> <jats:italic>ex</jats:italic> </jats:sub>), and the excess lifetime cancer risk (ELCR) were calculated as 145.45 Bq/kg, 122.13 nGy/h, 0.15 mSv/h, 0.39 and 0.29 × 10<jats:sup>−3</jats:sup> respectively.","PeriodicalId":21167,"journal":{"name":"Radiochimica Acta","volume":null,"pages":null},"PeriodicalIF":1.8,"publicationDate":"2023-12-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"138529146","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"化学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Saber Ibrahim, Ahmed M. Masoud, Mahmoud M. El-Maadawy, Hager Fahmy, Mohamed Taha
Plastic packaging waste is considered a serious threat to the environment due to its non-biodegradable nature. Transforming plastic waste into active carbons using pyrolysis methods could be a valuable option to solve the challenge of plastic waste. Synthesized active carbon was differentiated using zeta potential, particle size, SEM, BET, and DSC. This study also investigates the use of obtained active carbons for U(VI) removal from commercial phosphoric acid. The kinetics of adsorption were found to follow the pseudo-second-order model and intra-particle diffusion as one of the controlling mechanisms. Langmuir, and Freundlich, isotherms were employed to explore the equilibrium data. Furthermore, thermodynamic investigations revealed that uranium uptake is an endothermic, feasible, and spontaneous process. The present study concludes that plastic waste-based activated carbon could be employed as a low-cost alternative to commercial activated carbon for uranium removal from phosphoric acid and the production of green fertilizers.
{"title":"Recycling waste polymer packaging materials as effective active carbon porous materials for uranium removal from commercial phosphoric acid","authors":"Saber Ibrahim, Ahmed M. Masoud, Mahmoud M. El-Maadawy, Hager Fahmy, Mohamed Taha","doi":"10.1515/ract-2023-0165","DOIUrl":"https://doi.org/10.1515/ract-2023-0165","url":null,"abstract":"Plastic packaging waste is considered a serious threat to the environment due to its non-biodegradable nature. Transforming plastic waste into active carbons using pyrolysis methods could be a valuable option to solve the challenge of plastic waste. Synthesized active carbon was differentiated using zeta potential, particle size, SEM, BET, and DSC. This study also investigates the use of obtained active carbons for U(VI) removal from commercial phosphoric acid. The kinetics of adsorption were found to follow the pseudo-second-order model and intra-particle diffusion as one of the controlling mechanisms. Langmuir, and Freundlich, isotherms were employed to explore the equilibrium data. Furthermore, thermodynamic investigations revealed that uranium uptake is an endothermic, feasible, and spontaneous process. The present study concludes that plastic waste-based activated carbon could be employed as a low-cost alternative to commercial activated carbon for uranium removal from phosphoric acid and the production of green fertilizers.","PeriodicalId":21167,"journal":{"name":"Radiochimica Acta","volume":null,"pages":null},"PeriodicalIF":1.8,"publicationDate":"2023-11-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"138529110","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"化学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Mohamed Ragab Abass, Maha Ali Youssef, Marwa Ahmed Eid
This work is interested in the sorption and separation of 131Ba, 109Cd, 152+154Eu, and 97Zr from radioactive solutions onto barium molybdenum titanate loaded on carboxy methyl cellulose (BaMoTi@CMC) composites. In this work, different samples of BaMoTi@CMC composites were fabricated by the co-precipitation method and characterized using different analytical tools such as X-ray diffraction (XRD), attenuated total reflectance (ATR), and scanning electron microscope (SEM). The batch sorption investigations on 131Ba, 109Cd, 152+154Eu, and 97Zr include the influence of time, pH, and metal ion concentrations. The data reveal that S-3 has higher sorption efficiency than S-2 under all conditions. Isotherm is studied by Langmuir and Freundlich models. Binary systems data confirm that Cd(ii), Ba(ii), and Zr(iv) can be separated from Cd–Eu, Ba–Eu, and Zr–Eu binary systems using S-2 and S-3 at different pHs. Finally, the data prove that Zr(iv) and Ba(ii) can be easily separated from tertiary systems (Zr–Ba–Cd) onto S-2 and S-3 at pH 2.
{"title":"Inorganic composites based on carboxymethyl cellulose: preparation, characterization, sorption, and selectivity behavior for some radionuclides from radioactive solutions","authors":"Mohamed Ragab Abass, Maha Ali Youssef, Marwa Ahmed Eid","doi":"10.1515/ract-2023-0214","DOIUrl":"https://doi.org/10.1515/ract-2023-0214","url":null,"abstract":"This work is interested in the sorption and separation of <jats:sup>131</jats:sup>Ba, <jats:sup>109</jats:sup>Cd, <jats:sup>152+154</jats:sup>Eu, and <jats:sup>97</jats:sup>Zr from radioactive solutions onto barium molybdenum titanate loaded on carboxy methyl cellulose (BaMoTi@CMC) composites. In this work, different samples of BaMoTi@CMC composites were fabricated by the co-precipitation method and characterized using different analytical tools such as X-ray diffraction (XRD), attenuated total reflectance (ATR), and scanning electron microscope (SEM). The batch sorption investigations on <jats:sup>131</jats:sup>Ba, <jats:sup>109</jats:sup>Cd, <jats:sup>152+154</jats:sup>Eu, and <jats:sup>97</jats:sup>Zr include the influence of time, pH, and metal ion concentrations. The data reveal that S-3 has higher sorption efficiency than S-2 under all conditions. Isotherm is studied by Langmuir and Freundlich models. Binary systems data confirm that Cd(<jats:sc>ii</jats:sc>), Ba(<jats:sc>ii</jats:sc>), and Zr(<jats:sc>iv</jats:sc>) can be separated from Cd–Eu, Ba–Eu, and Zr–Eu binary systems using S-2 and S-3 at different pHs. Finally, the data prove that Zr(<jats:sc>iv</jats:sc>) and Ba(<jats:sc>ii</jats:sc>) can be easily separated from tertiary systems (Zr–Ba–Cd) onto S-2 and S-3 at pH 2.","PeriodicalId":21167,"journal":{"name":"Radiochimica Acta","volume":null,"pages":null},"PeriodicalIF":1.8,"publicationDate":"2023-11-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"138529138","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"化学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Rhodium-103m is one of the most attractive Auger electron emitters for internal radiotherapy. The half-life of 103mRh is relatively short (56.114 min). Therefore, it needs to be produced using a generator for clinical use. Most studies of 103Pd/103mRh generators using anion-exchange resins were carried out over 50 years ago. However, these resins are no longer commercially available. In the present study, we tested a 103Pd/103mRh generator using alternative anion-exchange resins (i.e., IRA904, IRA410, SA20A, and SA11AL). No-carrier-added 103Pd was used to make the generators. The 103mRh product was eluted from the generators using 6 mL of 0.1 M HCl with a flow rate 0.5 mL/min. The generator made from SA11AL showed good performance, with a yield of 39 %, an impurity level of 103Pd in the product of 0.29 %, and an operation time of 14 min. This makes this generator competitive with previously developed ones.
铑-103m是最具吸引力的内部放射治疗俄歇电子发射体之一。103mRh的半衰期相对较短(56.114 min)。因此，需要使用发生器生产用于临床。大多数使用阴离子交换树脂的103Pd/103mRh发电机的研究是在50多年前进行的。然而，这些树脂已不再在市场上销售。在本研究中，我们使用替代阴离子交换树脂(即IRA904, IRA410, SA20A和SA11AL)测试了103Pd/103mRh发生器。发生器采用无载流子添加的103Pd。103mRh产物用6ml 0.1 M HCl以0.5 mL/min的流速从发生器中洗脱。用SA11AL制备的发生器性能良好，产率为39%，产物中杂质含量为103Pd，为0.29%，运行时间为14 min，具有较好的竞争力。
{"title":"Production of Auger-electron-emitting 103mRh via a 103Pd/103mRh generator using an anion-exchange resin","authors":"Tomoyuki Ohya, Jun Ichinose, Kotaro Nagatsu, Yumi Sugo, Noriko Ishioka, Hiroshi Watabe, Masatoshi Itoh, Katsuyuki Minegishi, Ming-Rong Zhang","doi":"10.1515/ract-2023-0238","DOIUrl":"https://doi.org/10.1515/ract-2023-0238","url":null,"abstract":"Rhodium-103m is one of the most attractive Auger electron emitters for internal radiotherapy. The half-life of <jats:sup>103m</jats:sup>Rh is relatively short (56.114 min). Therefore, it needs to be produced using a generator for clinical use. Most studies of <jats:sup>103</jats:sup>Pd/<jats:sup>103m</jats:sup>Rh generators using anion-exchange resins were carried out over 50 years ago. However, these resins are no longer commercially available. In the present study, we tested a <jats:sup>103</jats:sup>Pd/<jats:sup>103m</jats:sup>Rh generator using alternative anion-exchange resins (i.e., IRA904, IRA410, SA20A, and SA11AL). No-carrier-added <jats:sup>103</jats:sup>Pd was used to make the generators. The <jats:sup>103m</jats:sup>Rh product was eluted from the generators using 6 mL of 0.1 M HCl with a flow rate 0.5 mL/min. The generator made from SA11AL showed good performance, with a yield of 39 %, an impurity level of <jats:sup>103</jats:sup>Pd in the product of 0.29 %, and an operation time of 14 min. This makes this generator competitive with previously developed ones.","PeriodicalId":21167,"journal":{"name":"Radiochimica Acta","volume":null,"pages":null},"PeriodicalIF":1.8,"publicationDate":"2023-11-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"138529149","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"化学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Effective treatment of uranium-containing wastewater is of great significance to the sustainable development of nuclear power and the protection of ecological environment. In this study, a highly efficient uranium adsorbent, graphene oxide (GO)/nano-hydroxyapatite (nHA) composite microspheres (nHA@rGO) was synthesized, which could effectively remove uranium from aqueous solution. Under the condition of pH = 3.5, T = 298 K, the maximum adsorption capacity reached 1672.96 mg/g. The results of batch experiments showed that the adsorption capacity of nHA@rGO microspheres was higher than that of nHA microspheres, indicating the enhancement of GO. The adsorption kinetics conformed to the pseudo second-order model. The changes of nHA@rGO microspheres before and after uranium adsorption were analyzed by FT-IR, XPS and XRD. The mechanisms of U(VI) ions adsorption onto nHA@rGO microspheres involved precipitation, surface complexation and ion exchange, in which the hydroxyl and phosphoric acid groups played important roles. The results showed that the prepared nHA@rGO microspheres can be used as an efficient and promising adsorbent for the treatment of uranium-containing wastewater.
有效处理含铀废水对核电的可持续发展和生态环境保护具有重要意义。本研究合成了一种高效的铀吸附剂——氧化石墨烯(GO)/纳米羟基磷灰石(nHA)复合微球(nHA@rGO)，可以有效地去除水溶液中的铀。在pH = 3.5, T = 298 K条件下，吸附量最大可达1672.96 mg/g。批量实验结果表明，nHA@rGO微球的吸附量高于nHA微球，表明氧化石墨烯的增强作用。吸附动力学符合准二级模型。利用FT-IR、XPS和XRD分析了nHA@rGO微球吸附铀前后的变化。nHA@rGO微球吸附U(VI)离子的机理包括沉淀、表面络合和离子交换，其中羟基和磷酸基团起重要作用。结果表明，制备的nHA@rGO微球可作为一种高效的含铀废水吸附剂。
{"title":"Efficient removal of U(VI) from aqueous solution by hydroxyapatite/graphene oxide composite microspheres","authors":"Wenjun Wu, Jianlong Wang","doi":"10.1515/ract-2023-0235","DOIUrl":"https://doi.org/10.1515/ract-2023-0235","url":null,"abstract":"Effective treatment of uranium-containing wastewater is of great significance to the sustainable development of nuclear power and the protection of ecological environment. In this study, a highly efficient uranium adsorbent, graphene oxide (GO)/nano-hydroxyapatite (nHA) composite microspheres (nHA@rGO) was synthesized, which could effectively remove uranium from aqueous solution. Under the condition of pH = 3.5, <jats:italic>T</jats:italic> = 298 K, the maximum adsorption capacity reached 1672.96 mg/g. The results of batch experiments showed that the adsorption capacity of nHA@rGO microspheres was higher than that of nHA microspheres, indicating the enhancement of GO. The adsorption kinetics conformed to the pseudo second-order model. The changes of nHA@rGO microspheres before and after uranium adsorption were analyzed by FT-IR, XPS and XRD. The mechanisms of U(VI) ions adsorption onto nHA@rGO microspheres involved precipitation, surface complexation and ion exchange, in which the hydroxyl and phosphoric acid groups played important roles. The results showed that the prepared nHA@rGO microspheres can be used as an efficient and promising adsorbent for the treatment of uranium-containing wastewater.","PeriodicalId":21167,"journal":{"name":"Radiochimica Acta","volume":null,"pages":null},"PeriodicalIF":1.8,"publicationDate":"2023-11-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"138529137","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"化学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Satyabrata Mishra, Pankaj, Chayan Patra, Debojyoti Ghosh, N. Desigan, P. Velavendan, K. A. Venkatesan, Ananthasivan Krishnamurthy
Abstract The aqueous waste generated during the treatment of Purex lean organic phase with alkaline carbonate solution contains washable degradation products and actinides in the form of carbonate complex. Management of such aqueous waste demands quantitative removal of the main degradation product, dibutyl phosphate (HDBP) and actinides from aqueous solution. In this context, batch studies have been carried out on the separation of HDBP from aqueous solution using n-dodecane (n-DD). For this purpose, the extraction behaviour of HDBP in n-DD was studied after acidification of alkaline carbonate solution with nitric acid. The studies with simulated waste containing HDBP and U(VI) nitrate showed the loss of U(VI) as precipitate during acidification as the uranium–DBP was poorly soluble in aqueous phase. However, the loss of U(VI) decreased with increase of aqueous phase acidity showing that the adjusted acidity of the carbonate waste plays an important role in the recovery of actinides.
{"title":"Studies on the acidification of carbonate waste stream for separation of di-butyl phosphate and recovery of metal","authors":"Satyabrata Mishra, Pankaj, Chayan Patra, Debojyoti Ghosh, N. Desigan, P. Velavendan, K. A. Venkatesan, Ananthasivan Krishnamurthy","doi":"10.1515/ract-2023-0196","DOIUrl":"https://doi.org/10.1515/ract-2023-0196","url":null,"abstract":"Abstract The aqueous waste generated during the treatment of Purex lean organic phase with alkaline carbonate solution contains washable degradation products and actinides in the form of carbonate complex. Management of such aqueous waste demands quantitative removal of the main degradation product, dibutyl phosphate (HDBP) and actinides from aqueous solution. In this context, batch studies have been carried out on the separation of HDBP from aqueous solution using n-dodecane (n-DD). For this purpose, the extraction behaviour of HDBP in n-DD was studied after acidification of alkaline carbonate solution with nitric acid. The studies with simulated waste containing HDBP and U(VI) nitrate showed the loss of U(VI) as precipitate during acidification as the uranium–DBP was poorly soluble in aqueous phase. However, the loss of U(VI) decreased with increase of aqueous phase acidity showing that the adjusted acidity of the carbonate waste plays an important role in the recovery of actinides.","PeriodicalId":21167,"journal":{"name":"Radiochimica Acta","volume":null,"pages":null},"PeriodicalIF":1.8,"publicationDate":"2023-11-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139267479","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"化学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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