Emily R. Mikeska, Natalie M. Lind, Alexander C. Ervin, Celine Khalife, Joseph P. Karnes, James D. Blakemore
Literature reports have demonstrated that Schiff-base-type ligands can serve as robust platforms for the synthesis of heterobimetallic complexes containing transition metals and the uranyl dication (UO22+). However, efforts have not advanced to include either synthesis of complexes containing second- or third-row transition metals or measurement of the redox properties of the corresponding heterobimetallic complexes, despite the significance of actinide redox in studies of nuclear fuel reprocessing and separations. Here, metalloligands denoted [Ni], [Pd], and [Pt] that contain the corresponding Group 10 metals have been prepared and a synthetic strategy to access species incorporating the uranyl ion (UO22+) has been explored, toward the goal of understanding how the secondary metals could tune uranium-centered redox chemistry. The synthesis and redox characterization of the bimetallic complex [Ni,UO2] was achieved, and factors that appear to govern extension of the chosen synthetic strategy to complexes with Pd and Pt are reported here. Infrared and solid-state structural data from X-ray diffraction analysis of the metalloligands [Pd] and [Pt] show that the metal centers in these complexes adopt the expected square planar geometries, while the structure of the bimetallic [Ni,UO2] reveals that the uranyl moiety influences the coordination environment of Ni(II), including inducement of a puckering of the ligand backbone of the complex in which the phenyl rings fold around the nickel-containing core in an umbrella-shaped fashion. Cyclic voltammetric data collected on the heterobimetallic complexes of both Ni(II) and Pd(II) provide evidence for uranium-centered redox cycling, as well as for the accessibility of other reductions that could be associated with Ni(II) or the organic ligand backbone. Taken together, these results highlight the unique redox behaviors that can be observed in multimetallic systems and design concepts that could be useful for accessing tunable multimetallic complexes containing the uranyl dication.
{"title":"Observations regarding the synthesis and redox chemistry of heterobimetallic uranyl complexes containing Group 10 metals","authors":"Emily R. Mikeska, Natalie M. Lind, Alexander C. Ervin, Celine Khalife, Joseph P. Karnes, James D. Blakemore","doi":"10.1515/ract-2023-0237","DOIUrl":"https://doi.org/10.1515/ract-2023-0237","url":null,"abstract":"Literature reports have demonstrated that Schiff-base-type ligands can serve as robust platforms for the synthesis of heterobimetallic complexes containing transition metals and the uranyl dication (UO<jats:sub>2</jats:sub> <jats:sup>2+</jats:sup>). However, efforts have not advanced to include either synthesis of complexes containing second- or third-row transition metals or measurement of the redox properties of the corresponding heterobimetallic complexes, despite the significance of actinide redox in studies of nuclear fuel reprocessing and separations. Here, metalloligands denoted [Ni], [Pd], and [Pt] that contain the corresponding Group 10 metals have been prepared and a synthetic strategy to access species incorporating the uranyl ion (UO<jats:sub>2</jats:sub> <jats:sup>2+</jats:sup>) has been explored, toward the goal of understanding how the secondary metals could tune uranium-centered redox chemistry. The synthesis and redox characterization of the bimetallic complex [Ni,UO<jats:sub>2</jats:sub>] was achieved, and factors that appear to govern extension of the chosen synthetic strategy to complexes with Pd and Pt are reported here. Infrared and solid-state structural data from X-ray diffraction analysis of the metalloligands [Pd] and [Pt] show that the metal centers in these complexes adopt the expected square planar geometries, while the structure of the bimetallic [Ni,UO<jats:sub>2</jats:sub>] reveals that the uranyl moiety influences the coordination environment of Ni(II), including inducement of a puckering of the ligand backbone of the complex in which the phenyl rings fold around the nickel-containing core in an umbrella-shaped fashion. Cyclic voltammetric data collected on the heterobimetallic complexes of both Ni(II) and Pd(II) provide evidence for uranium-centered redox cycling, as well as for the accessibility of other reductions that could be associated with Ni(II) or the organic ligand backbone. Taken together, these results highlight the unique redox behaviors that can be observed in multimetallic systems and design concepts that could be useful for accessing tunable multimetallic complexes containing the uranyl dication.","PeriodicalId":21167,"journal":{"name":"Radiochimica Acta","volume":"16 1","pages":""},"PeriodicalIF":1.8,"publicationDate":"2024-02-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139759124","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}
Hongtao Xia, Yuting Liu, Yang Wang, Zihao Feng, Qi Ren, Jianqi Lv, Yang Li, Yanjun Du, Yun Wang
An innovative phytic acid modified reed straw-derived hydrochar composite (PA-C-RBC) was prepared by using inexpensive reed straw and non-toxic phytic acid for the removal of uranium(VI) from aqueous environment. Several characterization results showed that PA-C-RBC was rough and porous with a large number of hydroxyl, carboxyl, and phosphate groups. The uranium(VI) adsorption process by PA-C-RBC conformed to pseudo-second-order kinetic and Langmuir models, and the theoretical maximal adsorption capacity could attain 418.78 mg/g at pH 5.0. PA-C-RBC had 72.66 % of selectivity and 6772.99 mL/g of distribution coefficient for U(VI). Due to the strong chelating between the hydroxyl and phosphate groups on PA-C-RBC and U(VI), PA-C-RBC had excellent adsorption selectivity. These finding highlighted a high potential for removing U(VI) from aqueous solutions.
{"title":"Incorporation of phytic acid into reed straw-derived hydrochar for highly efficient and selective adsorption of uranium(VI)","authors":"Hongtao Xia, Yuting Liu, Yang Wang, Zihao Feng, Qi Ren, Jianqi Lv, Yang Li, Yanjun Du, Yun Wang","doi":"10.1515/ract-2023-0250","DOIUrl":"https://doi.org/10.1515/ract-2023-0250","url":null,"abstract":"An innovative phytic acid modified reed straw-derived hydrochar composite (PA-C-RBC) was prepared by using inexpensive reed straw and non-toxic phytic acid for the removal of uranium(VI) from aqueous environment. Several characterization results showed that PA-C-RBC was rough and porous with a large number of hydroxyl, carboxyl, and phosphate groups. The uranium(VI) adsorption process by PA-C-RBC conformed to pseudo-second-order kinetic and Langmuir models, and the theoretical maximal adsorption capacity could attain 418.78 mg/g at pH 5.0. PA-C-RBC had 72.66 % of selectivity and 6772.99 mL/g of distribution coefficient for U(VI). Due to the strong chelating between the hydroxyl and phosphate groups on PA-C-RBC and U(VI), PA-C-RBC had excellent adsorption selectivity. These finding highlighted a high potential for removing U(VI) from aqueous solutions.","PeriodicalId":21167,"journal":{"name":"Radiochimica Acta","volume":"37 1","pages":""},"PeriodicalIF":1.8,"publicationDate":"2024-02-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139759130","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}
When operating and dismantling a nuclear facility that handles uranium, the surrounding soil may be contaminated, emphasizing the need for appropriate treatment and disposal methods for soil waste. This study assessed high-temperature sintering technology for uranium contaminated soil waste to overcome limitations in existing decontamination methods and the volume increase associated with current solidification technology. The sintering process was found to effectively vitrify and re-mineralize complex chemical components in the soil. Sintered bodies were produced under varying conditions, adjusting molding pressure, heating temperature, and time. Optimized conditions resulted in sintered bodies with a volume reduction rate exceeding 30 % and a compressive strength surpassing 10 MPa, indicating a significant impact on the phase conversion and re-mineralization of silt and clay minerals. The soil sintering mechanism was identified through comprehensive microscopic observations and mineral phase change analysis. Leaching evaluations of sintered bodies, made from simulated uranium-contaminated soil, demonstrated their applicability to contaminated soil wastes. Additionally, it was confirmed that the sintering temperature of the soil could be lowered by incorporating a small amount of B2O3, suggesting a means to enhance the economic feasibility of the treatment process. The findings of this study highlight the applicability of pressureless sintering technology, based on glass composite materials, capable of simultaneously reducing and stabilizing uranium-contaminated soil waste.
{"title":"Characterization of glass composite material by pressureless sintering of soil and its application to uranium contaminated soil as a waste form","authors":"Jaewoong Hwang, Jaseung Koo, Kenyoung Lee","doi":"10.1515/ract-2023-0222","DOIUrl":"https://doi.org/10.1515/ract-2023-0222","url":null,"abstract":"When operating and dismantling a nuclear facility that handles uranium, the surrounding soil may be contaminated, emphasizing the need for appropriate treatment and disposal methods for soil waste. This study assessed high-temperature sintering technology for uranium contaminated soil waste to overcome limitations in existing decontamination methods and the volume increase associated with current solidification technology. The sintering process was found to effectively vitrify and re-mineralize complex chemical components in the soil. Sintered bodies were produced under varying conditions, adjusting molding pressure, heating temperature, and time. Optimized conditions resulted in sintered bodies with a volume reduction rate exceeding 30 % and a compressive strength surpassing 10 MPa, indicating a significant impact on the phase conversion and re-mineralization of silt and clay minerals. The soil sintering mechanism was identified through comprehensive microscopic observations and mineral phase change analysis. Leaching evaluations of sintered bodies, made from simulated uranium-contaminated soil, demonstrated their applicability to contaminated soil wastes. Additionally, it was confirmed that the sintering temperature of the soil could be lowered by incorporating a small amount of B<jats:sub>2</jats:sub>O<jats:sub>3</jats:sub>, suggesting a means to enhance the economic feasibility of the treatment process. The findings of this study highlight the applicability of pressureless sintering technology, based on glass composite materials, capable of simultaneously reducing and stabilizing uranium-contaminated soil waste.","PeriodicalId":21167,"journal":{"name":"Radiochimica Acta","volume":"5 1","pages":""},"PeriodicalIF":1.8,"publicationDate":"2024-01-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139559238","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}
U. Tukhtaev, Shakhboz Khasanov, Jaloliddin Fayzullayev, A. Safarov, Bayramali Togaev, Seyedkarim Afsharipour
Abstract We conducted a comprehensive assessment of the Kattakurgan reservoir, alongside adjacent wells and boreholes, to measure the concentrations of natural radionuclides, heavy metals, and associated radiological hazards. Using NaI(Tl) crystal scintillation gamma spectrometers, we determined radionuclide levels in water and sediment. Inductively coupled plasma mass spectrometry (ICP-MS) was employed for heavy metal analysis. Our results showed radionuclide concentrations in reservoir water for 226Ra (0.8 Bq/L), 232Th (0.4 Bq/L), and 40K (0.4 Bq/L) were within the limits set by the World Health Organization (WHO). In contrast, deep well water samples showed elevated 226Ra concentrations (1.5 Bq/L). Sediment samples’ radionuclide levels were in line with UNSCEAR guidelines. Barium was the most notable heavy metal, with a concentration of 68.08 μg/L. While most radiation hazard indices remained within safety limits, the gamma index recorded a value of 1.057 Bq/kg. Our research provides valuable data for water quality assessment. The methods described can be applied to other reservoir studies. Regular monitoring is recommended for continuous safety evaluation, and further studies on biotic samples are suggested to enhance understanding of the reservoir’s ecosystem health.
{"title":"Determination of natural radionuclides and heavy metal concentrations in the groundwater and adjacent areas of the Kattakurgan reservoir, Uzbekistan","authors":"U. Tukhtaev, Shakhboz Khasanov, Jaloliddin Fayzullayev, A. Safarov, Bayramali Togaev, Seyedkarim Afsharipour","doi":"10.1515/ract-2023-0254","DOIUrl":"https://doi.org/10.1515/ract-2023-0254","url":null,"abstract":"Abstract We conducted a comprehensive assessment of the Kattakurgan reservoir, alongside adjacent wells and boreholes, to measure the concentrations of natural radionuclides, heavy metals, and associated radiological hazards. Using NaI(Tl) crystal scintillation gamma spectrometers, we determined radionuclide levels in water and sediment. Inductively coupled plasma mass spectrometry (ICP-MS) was employed for heavy metal analysis. Our results showed radionuclide concentrations in reservoir water for 226Ra (0.8 Bq/L), 232Th (0.4 Bq/L), and 40K (0.4 Bq/L) were within the limits set by the World Health Organization (WHO). In contrast, deep well water samples showed elevated 226Ra concentrations (1.5 Bq/L). Sediment samples’ radionuclide levels were in line with UNSCEAR guidelines. Barium was the most notable heavy metal, with a concentration of 68.08 μg/L. While most radiation hazard indices remained within safety limits, the gamma index recorded a value of 1.057 Bq/kg. Our research provides valuable data for water quality assessment. The methods described can be applied to other reservoir studies. Regular monitoring is recommended for continuous safety evaluation, and further studies on biotic samples are suggested to enhance understanding of the reservoir’s ecosystem health.","PeriodicalId":21167,"journal":{"name":"Radiochimica Acta","volume":"14 11","pages":""},"PeriodicalIF":1.8,"publicationDate":"2024-01-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139437313","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}
Abstract The current work deals with studying the influence of cheap, widespread CaO on improving the γ ray-protection capacity of the lightweight, sealing polyester. Therefore, polyester composites were filled with different concentrations of CaO fillers. The fabricated CaO-reinforced polyester composites’ density ranged between 1.177 g/cm3 and 1.377 g/cm3, when CaO filler concentrations increased between 0 wt% and 60 wt%, respectively. Fabricated composites’ morphology and chemical composition, as well as CaO fillers’ grain size and distribution, were proved using SEM and EDX spectroscopy. Additionally, the influence of CaO fillers on the gamma-ray shielding properties of the fabricated composites was evaluated using the Monte Carlo simulation and confirmed using the experimental measurements. The recorded results show an enhancement in the synthesized composites’ linear attenuation coefficient from 0.091 cm−1 to 0.106 cm−1 at a gamma ray energy of 0.662 MeV. Moreover, the excess in CaO concentration from 0 wt% and 60 wt% reduces the fabricated composites’ half-value thickness values from 7.64 cm to 6.51 cm, respectively.
摘要 当前的工作是研究廉价、广泛使用的氧化钙对提高轻质密封聚酯的γ射线防护能力的影响。因此，聚酯复合材料中填充了不同浓度的 CaO 填料。当 CaO 填料浓度在 0 wt% 和 60 wt% 之间增加时，制备的 CaO 增强聚酯复合材料的密度在 1.177 g/cm3 和 1.377 g/cm3 之间。利用扫描电镜和乙二胺四乙酸（EDX）光谱法证明了制备的复合材料的形貌和化学成分，以及 CaO 填料的粒度和分布。此外，还利用蒙特卡洛模拟评估了 CaO 填料对所制复合材料伽马射线屏蔽性能的影响，并通过实验测量进行了确认。记录结果显示，在伽马射线能量为 0.662 MeV 时，合成复合材料的线性衰减系数从 0.091 cm-1 提高到 0.106 cm-1。此外，氧化钙浓度从 0 wt% 增加到 60 wt%，复合材料的半值厚度值也从 7.64 cm 减小到 6.51 cm。
{"title":"CaO-enhanced polyester for safety: experimental study on fabrication, characterization, and gamma-ray attenuation","authors":"M. Marashdeh, K. A. Mahmoud","doi":"10.1515/ract-2023-0265","DOIUrl":"https://doi.org/10.1515/ract-2023-0265","url":null,"abstract":"Abstract The current work deals with studying the influence of cheap, widespread CaO on improving the γ ray-protection capacity of the lightweight, sealing polyester. Therefore, polyester composites were filled with different concentrations of CaO fillers. The fabricated CaO-reinforced polyester composites’ density ranged between 1.177 g/cm3 and 1.377 g/cm3, when CaO filler concentrations increased between 0 wt% and 60 wt%, respectively. Fabricated composites’ morphology and chemical composition, as well as CaO fillers’ grain size and distribution, were proved using SEM and EDX spectroscopy. Additionally, the influence of CaO fillers on the gamma-ray shielding properties of the fabricated composites was evaluated using the Monte Carlo simulation and confirmed using the experimental measurements. The recorded results show an enhancement in the synthesized composites’ linear attenuation coefficient from 0.091 cm−1 to 0.106 cm−1 at a gamma ray energy of 0.662 MeV. Moreover, the excess in CaO concentration from 0 wt% and 60 wt% reduces the fabricated composites’ half-value thickness values from 7.64 cm to 6.51 cm, respectively.","PeriodicalId":21167,"journal":{"name":"Radiochimica Acta","volume":"12 18","pages":""},"PeriodicalIF":1.8,"publicationDate":"2024-01-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139437728","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}
Abstract Acetonitrile is widely used as a solvent in synthesizing various fluorine-18 positron emission tomography (PET) radiotracers. Acetonitrile is classified as a Class II residual solvent, and due to its inherent toxic properties, the quantity of residual acetonitrile in drug products has to be limited. When working under Good Manufacturing Practices (GMP) during the radiosynthesis of a radiotracer, the aim is to control all solvent concentrations contained in the ready-to-use product. All products must meet predetermined specifications. Rarely, these limits may be exceeded. To avoid eliminating the entire batch, applying a straightforward time-based technique would be desirable to allow the majority of the product to be safely used. This technique should be based on determining a specific time and volume for which the radiotracer can be utilized in the patients after completing quality control analysis. Here, we report a very simple Excel sheet program based on existing mathematical equations that calculates the exact time and volume at which the radiotracer product can be safely administered to a patient.
{"title":"Overcoming the obstacle of excess acetonitrile content in the final fluorine-18 radiotracers","authors":"Mohammed Al-Qahtani","doi":"10.1515/ract-2023-0225","DOIUrl":"https://doi.org/10.1515/ract-2023-0225","url":null,"abstract":"Abstract Acetonitrile is widely used as a solvent in synthesizing various fluorine-18 positron emission tomography (PET) radiotracers. Acetonitrile is classified as a Class II residual solvent, and due to its inherent toxic properties, the quantity of residual acetonitrile in drug products has to be limited. When working under Good Manufacturing Practices (GMP) during the radiosynthesis of a radiotracer, the aim is to control all solvent concentrations contained in the ready-to-use product. All products must meet predetermined specifications. Rarely, these limits may be exceeded. To avoid eliminating the entire batch, applying a straightforward time-based technique would be desirable to allow the majority of the product to be safely used. This technique should be based on determining a specific time and volume for which the radiotracer can be utilized in the patients after completing quality control analysis. Here, we report a very simple Excel sheet program based on existing mathematical equations that calculates the exact time and volume at which the radiotracer product can be safely administered to a patient.","PeriodicalId":21167,"journal":{"name":"Radiochimica Acta","volume":"44 8","pages":""},"PeriodicalIF":1.8,"publicationDate":"2024-01-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139379866","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}
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":"44 1","pages":""},"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":"5 1","pages":""},"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":"110 1","pages":""},"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":"19 1","pages":""},"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}
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