Applied Radiation and Isotopes最新文献

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Radiation dose assessment and establishment of diagnostic reference levels for MDP Tc-99m bone scans: A multi-hospital study in Saudi Arabia

IF 1.6 3区工程技术 Q3 CHEMISTRY, INORGANIC & NUCLEAR

Applied Radiation and Isotopes

Pub Date : 2025-01-20 DOI: 10.1016/j.apradiso.2025.111695

Sultan Fahad Alhujaili , A. Sulieman , Ziyad Awadh Alrowaili , Abdulaziz Saad A. Alshabibi , Feras Alafer , Mansour M. Alqahtani , Hasan Salah , Abdullah J. Alanezi , Abdullah S. AlRashidi , Mohamed Abuzaid

Background

Bone scans using technetium-99m methylene diphosphonate (99mTc-MDP) are widely used in nuclear medicine for diagnosing various bone disorders. However, the radiation exposure associated with these procedures necessitates careful consideration and optimization.

Aim

This study aimed to evaluate radiation doses in terms of Effective Dose (ED) and establish Diagnostic Reference Levels (DRLs) for 99mTc-MDP bone scans in Saudi Arabia.

Method

Data were collected from 345 bone scan procedures across three hospitals in Saudi Arabia. Patient demographics, administered activities, and CT dose indicators were recorded. EDs were calculated using the RADAR Medical Procedure Radiation Dose Calculator. DRLs were established using the 75th percentile of administered activities.

Results

The study revealed variations in EDs across the three hospitals, with means ranging from 4.14 to 5.46 mSv for 99mTc-MDP bone scans. CT dose indicators were generally lower than those reported in some international studies. The established DRLs (75th percentile) for administered activities were 771 MBq, 841 MBq, and 1006.4 MBq for the three hospitals, respectively. These values were within the range of international DRLs but showed some variation compared to other Saudi Arabian studies.

Conclusion

This study provides valuable insights into current practices for 99mTc-MDP bone scans in Saudi Arabia. The observed variations in doses and established DRLs highlight the need for standardization and optimization of protocols across hospitals. These findings can serve as a benchmark for nuclear medicine facilities in Saudi Arabia, promoting radiation safety while maintaining diagnostic quality.

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引用次数: 0

Investigating the Li2O-Al2O3-B2O3 matrix: A promising approach for immobilizing radioactive waste

IF 1.6 3区工程技术 Q3 CHEMISTRY, INORGANIC & NUCLEAR

Applied Radiation and Isotopes

Pub Date : 2025-01-20 DOI: 10.1016/j.apradiso.2025.111696

Syaza Amira Zulkeplee, Nor Ezzaty Ahmad, Mohamad Syazwan Mohd Sanusi, Suhairul Hashim

Dealing with radioactive waste, particularly from various industrial processes, poses significant challenges. This paper explores the use of lithium aluminate borate (Li-Al-B) glass matrix as an alternative method for immobilizing radioactive waste, focusing specifically on waste generated in tin smelting industries, known as tin slag. The study primarily concentrates on transforming tin slag, a byproduct abundant in Natural Occurring Radioactive Material (NORM), into a stable and safe form for disposal. The experimental procedures involve blending different compositions of tin slag and Li-Al-B glass, followed by melting them at 1000 °C for 1 h and then rapidly cooling to room temperature. The resulting glass waste identifies an optimal weight percentage of waste loading (typically ranging from 25% to 45%), to minimize volume while effectively immobilizing radioactive material. Notably, the glass waste exhibited an amorphous phase during the product consistency test (PCT) process, demonstrating the fundamental relationship between waste composition and immobilization efficiency. Energy dispersive X-ray spectroscopy (EDX) analysis confirmed a uniform distribution of major elements within the glass waste, underscoring its structural integrity. Furthermore, the dissolution rate of key elements in the glass waste is analyzed, revealing a robust resistance to leaching under varying pH conditions. The normalized mass loss of Boron (B), Lithium (Li), and Aluminum (Al) consistently remain below established glass limits (<2 gm^-2), indicative of the glass's exceptional durability. In conclusion, these findings highlight the potential effectiveness of Li-Al-B glass as a versatile host material for immobilizing solid radioactive waste, extending beyond its initial application with tin slag. By highlighting the positive qualities of this matrix, the study emphasizes its potential flexibility in accommodating various types of solid waste matrices.

{"title":"Investigating the Li2O-Al2O3-B2O3 matrix: A promising approach for immobilizing radioactive waste","authors":"Syaza Amira Zulkeplee, Nor Ezzaty Ahmad, Mohamad Syazwan Mohd Sanusi, Suhairul Hashim","doi":"10.1016/j.apradiso.2025.111696","DOIUrl":"10.1016/j.apradiso.2025.111696","url":null,"abstract":"<div><div>Dealing with radioactive waste, particularly from various industrial processes, poses significant challenges. This paper explores the use of lithium aluminate borate (Li-Al-B) glass matrix as an alternative method for immobilizing radioactive waste, focusing specifically on waste generated in tin smelting industries, known as tin slag. The study primarily concentrates on transforming tin slag, a byproduct abundant in Natural Occurring Radioactive Material (NORM), into a stable and safe form for disposal. The experimental procedures involve blending different compositions of tin slag and Li-Al-B glass, followed by melting them at 1000 °C for 1 h and then rapidly cooling to room temperature. The resulting glass waste identifies an optimal weight percentage of waste loading (typically ranging from 25% to 45%), to minimize volume while effectively immobilizing radioactive material. Notably, the glass waste exhibited an amorphous phase during the product consistency test (PCT) process, demonstrating the fundamental relationship between waste composition and immobilization efficiency. Energy dispersive X-ray spectroscopy (EDX) analysis confirmed a uniform distribution of major elements within the glass waste, underscoring its structural integrity. Furthermore, the dissolution rate of key elements in the glass waste is analyzed, revealing a robust resistance to leaching under varying pH conditions. The normalized mass loss of Boron (B), Lithium (Li), and Aluminum (Al) consistently remain below established glass limits (<2 gm<sup>-2</sup>), indicative of the glass's exceptional durability. In conclusion, these findings highlight the potential effectiveness of Li-Al-B glass as a versatile host material for immobilizing solid radioactive waste, extending beyond its initial application with tin slag. By highlighting the positive qualities of this matrix, the study emphasizes its potential flexibility in accommodating various types of solid waste matrices.</div></div>","PeriodicalId":8096,"journal":{"name":"Applied Radiation and Isotopes","volume":"218 ","pages":"Article 111696"},"PeriodicalIF":1.6,"publicationDate":"2025-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143051416","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

99mTc radiolabeling of palm shell charcoal: A preliminary study for potential lung ventilation scintigraphy agent

IF 1.6 3区工程技术 Q3 CHEMISTRY, INORGANIC & NUCLEAR

Applied Radiation and Isotopes

Pub Date : 2025-01-19 DOI: 10.1016/j.apradiso.2025.111689

I. Saptiama , M. Munir , I.S. Qaira , F. Rindiyantono , A. Nurmanjaya , K.E. Prasetya , M. Subechi , Marlina , Abidin , G. Suhariyono , Ferry , K.S.S. Putri , A.S. Wismogroho , C. Firdharini

To investigate the potential of activated carbon from palm kernel shell waste for ^99mTc-radiolabeled nanocarbon aerosol, a new production technology for carbon-based ^99mTc-radioaerosol from such a waste was developed. Treated-palm shell charcoal (t-PSC) was prepared by hydrothermal method to increase the surface area, followed by ^99mTc radiolabelling optimization. The optimal ^99mTc radiolabeling conditions resulted in an adsorption capacity of 21.43 ng Re/g t-PSC (8.32 GBq ^99mTc/g t-PSC). After high-energy milling treatment, fines particle fraction (FPF), and median mass aerodynamic diameter (MMAD) of the milled t-PSC were 28.34 ± 0.61%, and 8.31 ± 2.31 μm, respectively. The results imply that ^99mTc-labeled t-PSC has a potential for lung ventilation scan agents with the optimization of milling process to reduce the aerodynamic size within the optimal lung delivery of less than 5 μm.

{"title":"99mTc radiolabeling of palm shell charcoal: A preliminary study for potential lung ventilation scintigraphy agent","authors":"I. Saptiama , M. Munir , I.S. Qaira , F. Rindiyantono , A. Nurmanjaya , K.E. Prasetya , M. Subechi , Marlina , Abidin , G. Suhariyono , Ferry , K.S.S. Putri , A.S. Wismogroho , C. Firdharini","doi":"10.1016/j.apradiso.2025.111689","DOIUrl":"10.1016/j.apradiso.2025.111689","url":null,"abstract":"<div><div>To investigate the potential of activated carbon from palm kernel shell waste for <sup>99m</sup>Tc-radiolabeled nanocarbon aerosol, a new production technology for carbon-based <sup>99m</sup>Tc-radioaerosol from such a waste was developed. Treated-palm shell charcoal (t-PSC) was prepared by hydrothermal method to increase the surface area, followed by <sup>99m</sup>Tc radiolabelling optimization. The optimal <sup>99m</sup>Tc radiolabeling conditions resulted in an adsorption capacity of 21.43 ng Re/g t-PSC (8.32 GBq <sup>99m</sup>Tc/g t-PSC). After high-energy milling treatment, fines particle fraction (FPF), and median mass aerodynamic diameter (MMAD) of the milled t-PSC were 28.34 ± 0.61%, and 8.31 ± 2.31 μm, respectively. The results imply that <sup>99m</sup>Tc-labeled t-PSC has a potential for lung ventilation scan agents with the optimization of milling process to reduce the aerodynamic size within the optimal lung delivery of less than 5 μm.</div></div>","PeriodicalId":8096,"journal":{"name":"Applied Radiation and Isotopes","volume":"218 ","pages":"Article 111689"},"PeriodicalIF":1.6,"publicationDate":"2025-01-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143045610","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Adsorption efficiency correction for 226Ra determination on drinking water using Ra-Nucfilm disc

IF 1.6 3区工程技术 Q3 CHEMISTRY, INORGANIC & NUCLEAR

Applied Radiation and Isotopes

Pub Date : 2025-01-19 DOI: 10.1016/j.apradiso.2025.111690

Héloïse Gervot , Christine Germond , Jérome Loess , Dominique Saurat

A specific goal of the French army is to ensure health protection of soldiers consuming water from overseas operations territories. To do so, the French Defence Radiological Protection Service developed a method to quantify the amount of ²²⁶Ra in water using Ra-Nucfilm discs. ²²⁶Ra analysis is achieved within less than two days and the detection limit, function of adsorption efficiency, ranges from 0.85 to 25 mBq/L. The novelty is the determination of a factor correcting adsorption efficiency, equal to 1.3, from the difference in adsorption between ²²⁶Ra and ¹³³Ba using as internal tracer.

引用次数: 0

Recovered water - H218O from the 18F[FDG] production as liquid radioactive waste 从 18F[FDG] 生产中回收的水 - H218O 作为液态放射性废物。

IF 1.6 3区工程技术 Q3 CHEMISTRY, INORGANIC & NUCLEAR

Applied Radiation and Isotopes

Pub Date : 2025-01-19 DOI: 10.1016/j.apradiso.2025.111691

Magdalena Długosz-Lisiecka , Teresa Jakubowska

In this study, ten recovered water samples were analysed using gamma spectrometry and Liquid Scintillation Counting techniques for identification of radioactive impurities (quality and quantity) and for radioactive waste qualifications. The presence of several radioactive isotopes of ³H, ^56,57Co ⁵²Mn in the recovered [¹⁸O] water irradiated with 11 MeV protons used to produce [¹⁸F] fluoride by the ¹⁸O(p,n)¹⁸F reaction has been confirmed. Radioactive impurities were generated directly in enriched water or washed out from activated Havar foil, or tantalum body target material. The highest impact on the qualification of the recovered water remains after the production as a radioactive waste has ⁵⁶Co. The highest activity concentration of about 0.1 GBq/ml has been detected in the case of tritium ³H. All ten samples were qualified as transitional, low-level radioactive wastes.

引用次数: 0

Development of a current measurement system for 4πγ ionization chamber

IF 1.6 3区工程技术 Q3 CHEMISTRY, INORGANIC & NUCLEAR

Applied Radiation and Isotopes

Pub Date : 2025-01-18 DOI: 10.1016/j.apradiso.2025.111684

Cong Chen , Weihua Zeng , Zhijie Yang , Shengli Hou , Juncheng Liang , Haoran Liu

A current measurement system named ICMS-TIA (Ionization Current Measurement System based on Transimpedance Amplifier) for rapid and accurate measurement of the ionization current of 4πγ ionization chamber is developed based on an improved high-value resistance I-V conversion method with range switching function. The effective measurement range of the ICMS-TIA is from 50 fA to 50 μA and the response time of the system is less than 2.8 s. The test shows that the deviation of the current measurement above 1 pA is less than 0.41%, and the uncertainty over a range of current levels is given. In the linearity tests, the maximum deviation is 1.15 %. These tests indicate that this system can be used for current measurement of 4πγ ionization chamber.

引用次数: 0

Gravimetric drop-on-demand inkjet deposition for imaging phantoms with traceable activity

IF 1.6 3区工程技术 Q3 CHEMISTRY, INORGANIC & NUCLEAR

Applied Radiation and Isotopes

Pub Date : 2025-01-18 DOI: 10.1016/j.apradiso.2025.111683

Sean P. Jollota , Denis E. Bergeron , R. Michael Verkouteren , Svetlana Nour , Ryan P. Fitzgerald , Ohyun Kwon , Jeffrey L. Radtke , Bryan P. Bednarz , Brian W. Miller , Larry A. DeWerd

Autoradiography is an important tool in preclinical investigations with alpha-emitting radionuclides. Precise biodistribution studies establish the effectiveness of biological “targeting” strategies, assuring that the promise of targeted alpha therapy (TAT) is realized. There is a need for calibration and test objects (phantoms) to evaluate and support precise autoradiography. Leveraging recent advances in drop-on-demand inkjet methods, we demonstrate the preparation of an autoradiography phantom with activity traceable to primary measurements. With digital autoradiography we demonstrate spatial resolution less than 20 μm. Corrections required to calibrate a digital autoradiography system for activity are discussed.

引用次数: 0

Integral cross section measurement of the 19F( γ, n)18F reaction for 12-20 MeV bremsstrahlung photons 12-20 MeV轫致辐射光子19F(γ, n)18F反应的积分截面测量。

IF 1.6 3区工程技术 Q3 CHEMISTRY, INORGANIC & NUCLEAR

Applied Radiation and Isotopes

Pub Date : 2025-01-18 DOI: 10.1016/j.apradiso.2025.111673

Shaima Akbar , M.M. Musthafa , C.V. Midhun , Swapna Lilly Cyriac , Resmi K. Bharathan , Silpa Ajaykumar , Antony Joseph , K.C. Jagadeesan

¹⁸F radioactive isotope is widely used in PET imaging for nuclear medicine. Medical linear accelerators producing high-flux bremsstrahlung beams up to 20 MeV are commonly used in radiation therapy. Hence, the production of ¹⁸F through photon-induced channels will reduce many of the intricacies in transportation and handling. With this objective, the integral cross sections for ¹⁹F(

γ

, n)¹⁸F reaction, for bremsstrahlung endpoint energies of 12, 14.6, and 20 MeV, are measured employing the activation technique. The experimentally measured cross section was analysed using the nuclear reaction code TALYS 1.96. Parameters for level density models and gamma strength function models are optimized within the framework of a statistical approach.

18F放射性同位素广泛应用于核医学PET成像。医用线性加速器产生高达20兆电子伏特的高通量轫致辐射光束，常用于放射治疗。因此，通过光子诱导通道生产18F将减少运输和处理中的许多复杂性。为此，采用活化技术测量了19F(γ, n)18F反应在轫致辐射端点能量为12、14.6和20 MeV时的积分截面。实验测量的截面用核反应代码TALYS 1.96进行分析。在统计方法的框架内优化了水平密度模型和伽马强度函数模型的参数。

引用次数: 0

Assessment of the effectiveness of Monte Carlo Simulation for dose control at the Moroccan Boukhalef ionization facility

IF 1.6 3区工程技术 Q3 CHEMISTRY, INORGANIC & NUCLEAR

Applied Radiation and Isotopes

Pub Date : 2025-01-17 DOI: 10.1016/j.apradiso.2025.111682

Youssef El-ouardi , Adil Aknouch , Halima Jemmal

Controlling the absorbed dose received by a target is a major challenge encountered during ionizing radiation applications. For experimentally measuring absorbed dose, dosimetric systems are used. On the other hand, in addition to experimental methods of dose measurement, there are other alternatives for calculating absorbed doses, these are numerical methods based on the Monte Carlo method which are very sophisticated and widely used throughout the world. The objective of this work is to demonstrate the effectiveness of Monte Carlo simulation for the contribution to dose control during the dosimetry operation at the Moroccan Boukhalef ionization facility (SIBO). In this study, a comparison is made between Monte Carlo simulation and a dose measurement experiment through the EPR/alanine dosimetry system. This comparison is made in terms of absorbed dose rate in a cylindrical shaped container filled with Expanded Polystyrene (EPS). The results obtained show a good agreement between the Monte simulation and the dose measurement experiment with differences not exceeding 6%. Therefore, Monte Carlo simulation is able to replace or complement experimental methods of dose measurements at the Moroccan Bokhlalef ionization facility.

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引用次数: 0

Radiation spectroscopy of irradiated VVER-1200 fuel with burnable Am-absorber: A computational approach

IF 1.6 3区工程技术 Q3 CHEMISTRY, INORGANIC & NUCLEAR

Applied Radiation and Isotopes

Pub Date : 2025-01-16 DOI: 10.1016/j.apradiso.2025.111674

Sergey V. Bedenko , Gennady N. Vlaskin , Sergey D. Polozkov , Dmitry G. Veretennikov , Alexey S. Demin , Nima Ghal-Eh , Faezeh Rahmani

Safe storage of fresh and irradiated fuel is ensured by solving the problem of photon emission protection. The neutron component is usually not taken into account due to its low intensity. However, for the new VVER-1200 fuel, the neutron component consideration is a mandatory procedure for radiation safety. In this study, the radiation dose was calculated for a fuel consisting of UO₂ with a heterogeneous distribution of Am₂O₂ microcapsules, and the (α, n) component of the neutron background was evaluated. A comparative analysis of radiation characteristics of fuel assemblies shows that there is a significant excess in both the neutron and the photon components of the fuel under study. The yield and dose of neutrons from Am−containing fuel exceed those of uranium−based fuels by a factor of two, and when calculating the dose, it is necessary to take into account the energy spectrum of (α, n) neutrons in Am₂O₂ microcapsules. The analysis of the impact of Am on the photon component indicates that ensuring radiation safety for both fresh and irradiated fuels necessitates addressing challenges associated with photon radiation protection. This study aims to establish comprehensive procedures and guidelines for the handling of novel fuel materials during both production and post-irradiation processes within the reactor environment.

{"title":"Radiation spectroscopy of irradiated VVER-1200 fuel with burnable Am-absorber: A computational approach","authors":"Sergey V. Bedenko , Gennady N. Vlaskin , Sergey D. Polozkov , Dmitry G. Veretennikov , Alexey S. Demin , Nima Ghal-Eh , Faezeh Rahmani","doi":"10.1016/j.apradiso.2025.111674","DOIUrl":"10.1016/j.apradiso.2025.111674","url":null,"abstract":"<div><div>Safe storage of fresh and irradiated fuel is ensured by solving the problem of photon emission protection. The neutron component is usually not taken into account due to its low intensity. However, for the new VVER-1200 fuel, the neutron component consideration is a mandatory procedure for radiation safety. In this study, the radiation dose was calculated for a fuel consisting of UO<sub>2</sub> with a heterogeneous distribution of Am<sub>2</sub>O<sub>2</sub> microcapsules, and the (α, n) component of the neutron background was evaluated. A comparative analysis of radiation characteristics of fuel assemblies shows that there is a significant excess in both the neutron and the photon components of the fuel under study. The yield and dose of neutrons from Am−containing fuel exceed those of uranium−based fuels by a factor of two, and when calculating the dose, it is necessary to take into account the energy spectrum of (α, n) neutrons in Am<sub>2</sub>O<sub>2</sub> microcapsules. The analysis of the impact of Am on the photon component indicates that ensuring radiation safety for both fresh and irradiated fuels necessitates addressing challenges associated with photon radiation protection. This study aims to establish comprehensive procedures and guidelines for the handling of novel fuel materials during both production and post-irradiation processes within the reactor environment.</div></div>","PeriodicalId":8096,"journal":{"name":"Applied Radiation and Isotopes","volume":"218 ","pages":"Article 111674"},"PeriodicalIF":1.6,"publicationDate":"2025-01-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143021831","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

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