Subjects: Chemistry >> Nuclear Chemistry submitted time 2024-07-22
Abstract： Technetium plays an important role in the vitrification of high level radioactive waste (HLW) and the geological disposal of high level radioactive waste (HLW) in the latter part of the nuclear fuel cycle. In this work, an optimized process for extracting technetium from post-treatment nitric acid medium with NTAamide (C8) as extractant was proposed.Based on the principle of NIAamide (C8) to extract technetium, reverse stem technetium ammonium carbonate and oxalic acid to wash impurity ions, a process for extracting and purifying technetium from post-treatment tailings was designed. The results show that the recovery of technetium is 99.9%, and the purification coefficients of Sr, Cs, Zr and Naphthalene in technetium are 6.9×103, 7.9×104, 4.3×102 and 45 respectively.
Subjects: Chemistry >> Nuclear Chemistry submitted time 2024-07-11
Abstract： Background: Medical radioactive isotope 99mTc is widely used in SPECT imaging, primarily derived from the decay of its parent nucleus 99Mo. With 90% of the global 99Mo supply coming from five reactors that will be decommissioned, there is a pressing need for a stable supply. Using proton accelerators to irradiate 100Mo to produce 99Mo and 99mTc has become an important supplementary method. Some researchers have used Monte Carlo software FLUKA to simulate the yield of radioactive isotopes, discovering discrepancies between the simulation results and experimental measurements. Purpose: Verification of whether FLUKA can accurately simulate the production of 99Mo and 99mTc from proton irradiation of 100Mo. Method: This paper calculates the activation function used by FLUKA and compares the results with those from other databases and nuclear reaction programs such as TALYS. A combined method using FLUKA and other databases' activation function calculations is proposed to estimate the yield of 99Mo and 99mTc after proton irradiation and cooling. Results: FLUKA underestimates the yield of 99Mo by approximately 59%-67% and overestimates the yield of 99mTc by approximately 245%-260% when compared to TENDL2023 activation function calculations. Conclusions: The discrepancies arise from the significant differences between the proton-nucleus reaction cross-sections in FLUKA and those in other databases, and FLUKA's inability to accurately handle the production of different excited states of nuclides.
Subjects: Chemistry >> Nuclear Chemistry submitted time 2024-07-08
Abstract： Plastic scintillation microspheres (PSm) are a novel luminescent material with significant potential in the measurement and continuous monitoring of radionuclide activity. [Purpose]: To optimize the application of PSm in radionuclide activity analysis, it is crucial to understand the impact of different measurement conditions on the background count rate and counting efficiency of PSm. [Methods]: The chemiluminescence and photoluminescence of PSm were measured using a Hidex 300SL liquid scintillation counter. The study also investigated the effect of PSm dosage and the type of liquid scintillation vial on the background count rate of PSm. Using 14C as a reference radionuclide, the counting efficiency of PSm under different conditions was determined. Finally, a 20 mL plastic liquid scintillation vial with the lowest detection limit was selected for the activity analysis of 14C solution. [Results]: Weak interference signals from chemiluminescence and photoluminescence were observed when PSm was mixed with water. This interference can be mitigated by by keeping the samples in a light-protected environment. The background count rate was positively correlated with the volume of PSm, with larger volumes resulting in higher background counts. The counting efficiency was positively correlated with the height of PSm, with higher heights corresponding to higher counting efficiency. Differences between plastic and glass liquid scintillation vials of comparable specifications were minimal. The minimum detectable activity (MDA) of the 20 mL vial was lower. Analysis of 14C solutions in the range of 574.5 to 3825.6Bq·L-1 showed good agreement between measured and expected values. [Conclusions]: The study demonstrates that the background count rate and counting efficiency of plastic scintillation microspheres are influenced by various factors such as sample volume. Under fixed conditions, PSm measurements exhibit good accuracy. PSm shows promising potential in radionuclide analysis.
Subjects: Chemistry >> Nuclear Chemistry submitted time 2024-06-18
Abstract： As an advanced radiation therapy technique, proton therapy is valued for its precise dose distribution and minimized damage to surrounding normal tissues. However, accurate dosimetry of proton beams is crucial to ensure treatment efficacy and safety. Fiber optic dosimeters have become a hotspot for proton therapy dosimetry research due to their unique advantages, such as high spatial resolution, real-time monitoring capability, water-equivalence, and resistance to electromagnetic interference. Based on this, this paper reviews the technical principles, current application status, development challenges and future trends of fiber optic dosimeters in proton therapy dosimetry. Firstly, this paper outlines the luminescence mechanisms of fiber optic dosimeters and describes their working principles. Then, the paper discusses the utility of fiber optic dosimeters based on different luminescence mechanisms in clinical proton therapy for high dose rate adaptability and accuracy. Meanwhile, the paper also points out the challenges and solutions to the current fiber optic dosimeter technology, such as applying scintillation quenching correction, improving time responsiveness, optimizing the data acquisition system, and enhancing the correction capability for interfering factors. Finally, the paper looks at the future of fiber optic dosimeter technology, especially its potential application in ultra-high dose rate treatments such as Flash therapy, and how existing limitations can be overcome through innovations in materials science and optoelectronic technology. Through this review, it aims to provide researchers and clinicians with a comprehensive perspective to improve the accuracy and therapeutic efficacy of proton therapy.
Subjects: Chemistry >> Nuclear Chemistry submitted time 2024-06-12
Abstract： Background: The Ir-192 brachytherapy source is a high dose rate gamma radiation source characterized by a high central dose rate and a rapid dose fall-off at the periphery. In clinical treatments, this dose distribution allows the Ir-192 source to effectively protect the normal tissues and organs surrounding the tumor. Objective To establish a detailed structural model of a domestically produced high-dose-rate Ir-192 brachytherapy source using Monte Carlo simulation software, based on the dosimetric parameters recommended by the American Association of Physicists in Medicine (AAPM) in TG43-U1, and to perform simulation calculations. Methods Using the Monte Carlo software, a detailed structural model of the domestically produced high-dose-rate Ir-192 brachytherapy source was established. The simulated dosimetric parameters included the dose rate constant Λ, air kerma rate per unit activity, radial dose function , and anisotropy function . Results The simulated dose rate constant was 1.105 cGy·h-1·U-1, with a difference of less than 1.2% from the literature values. Air kerma rate per unit activity was 9.788×10-8 UBq-1, with a difference of 0.23% from the literature values. The radial dose function was obtained for distances from 0.5 to 20 cm from the source axis, and an empirical formula was fitted. Conclusions The domestically produced Ir-192 source model established using the Monte Carlo software shows good consistency with the literature-reported dosimetric parameters, indicating that this model can be used for clinical practice applications of domestically produced Ir-192 sources and has certain guiding significance.
Subjects: Chemistry >> Nuclear Chemistry submitted time 2024-06-07
Abstract： Background: The Ir-192 brachytherapy source is a high dose rate gamma radiation source characterized by a high central dose rate and a rapid dose fall-off at the periphery. In clinical treatments, this dose distribution allows the Ir-192 source to effectively protect the normal tissues and organs surrounding the tumor. Objective To establish a detailed structural model of a domestically produced high-dose-rate Ir-192 brachytherapy source using Monte Carlo simulation software, based on the dosimetric parameters recommended by the American Association of Physicists in Medicine (AAPM) in TG43-U1, and to perform simulation calculations. Methods Using the Monte Carlo software, a detailed structural model of the domestically produced high-dose-rate Ir-192 brachytherapy source was established. The simulated dosimetric parameters included the dose rate constant Λ, air kerma rate per unit activity, radial dose function , and anisotropy function . Results The simulated dose rate constant was 1.105 cGy·h-1·U-1, with a difference of less than 1.2% from the literature values. Air kerma rate per unit activity was 9.788×10-8 UBq-1, with a difference of 0.23% from the literature values. The radial dose function was obtained for distances from 0.5 to 20 cm from the source axis, and an empirical formula was fitted. Conclusions The domestically produced Ir-192 source model established using the Monte Carlo software shows good consistency with the literature-reported dosimetric parameters, indicating that this model can be used for clinical practice applications of domestically produced Ir-192 sources and has certain guiding significance.
Subjects: Chemistry >> Nuclear Chemistry submitted time 2024-05-02
Abstract： Abstract Background：Human epidermal growth factor receptor 2 (HER2) is widely present in many malignant tumors.It is associated with poor prognosis. Particularly in breast cancer, however, there is heterogeneity in HER2 expression.Currently immunohistochemistry and fluorescence
are used for assessing HER2 status. However they have so much significant limitations.HER2 receptor imaging has significant advantages. It is a potential option for detecting HER2-positive lesions of Radiolabeled mimetic peptides.Purpose：To prepare a 18F labeled human epidermal growth factor receptor2(HER2) peptide B2-S22-AFA(18F-NFP-TP1296),and evaluate its biodistribution and microPET characteristics.Methods:The tracer conjugate was labeled with 18F in one step.18F-NFP-TP1296 was performed in vitro studies and MicroPET imaging in the SKBR-3 breast cancer model.Rusults:18F-NFP-TP1296 was synthesized in about 30 min with the non-decay corrected yields more than 10%, and radiochemical purity more than 95%.MicroPET imaging revealed that the SKBR-3 xenografts were visualized and the tumor uptakes were 5.63±0.14%ID/g、6.26±0.27%ID/g and 5.83±0.44%ID/g at 30min、60min and 120min.The corresponding tumor-to-blood and tumor-to-muscle ratios were 3.21±0.32、4.08±0.73 and 1.69±0.18 respectively, and 1.55±0.11%ID/g、1.84±0.12%ID/g and 3.10±0.30%ID/g at 30、60 and 120min. The lung metastasis tumor uptakes were 2.2%ID/g、2.5%ID/g and 2.1%ID/g at 30、60 and 120min.Conclusion：18F-NFP-TP1296 can be successfully labeled by one-step method.The 18F-NFP-TP1296 probe owns the advantages of favorable imaging properties, convenient preparation, excellent stability, safety, rapid clearance in the blood, which support its application for further research.
Subjects: Chemistry >> Nuclear Chemistry submitted time 2024-04-09
Abstract： In the process of accelerated development of nuclear medicine department in recent years, the construction of decay tanks and the storage time of radioactive wastewater containing I-131 have become issues of great concern for environmental regulatory agencies and hospitals. Basic Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources (GB 18871), Radiation Protection and Safety Requirements for Nuclear Medicine (HJ 1188), and Reply to Consultation on Several Clauses of the Nuclear Medicine Standard have stipulated the compliant discharge methods for radioactive wastewater containing I-131 from hospitals. This paper presents a theoretical calculation formula for the total activity of iodine-131 when a delay tank in a hospital is full. It demonstrates that among the three compliant disperse methods for radioactive wastewater containing iodine-131 in the decay tank, the method specified in GB 18871 is advantageous for the operation of the nuclear medicine department in the hospital. The paper also introduces the RJ equation group, which addresses the calculation of minimum decay time and volume of the decay tank. The actual measured data from four hospitals demonstrates that when the temporary storage period for radioactive wastewater containing iodine-131 reaches the minimum time calculated by the RJ equations, the total discharge activity of iodine-131 complies with the national environmental protection standards.These findings provide clear and specific guidance for the construction of decay tanks in nuclear medicine departments and for the supervision and inspection conducted by regulatory authorities.
Subjects: Chemistry >> Nuclear Chemistry submitted time 2023-06-13
Abstract： Objective: We herein fabricated a branched structure containing AO groups on polypropylene/polyethylene spun-laced nonwoven (PP/PE SNW) fibers using grafting polymerization induced by radiation (RIGP) to improve AO utilization. Methods: The chemical structures, thermal properties, and surface morphologies of the raw and treated PP/PE SNW fibers were studied. The adsorption properties were investigated using batch adsorption experiments in simulated seawater with an initial uranium concentration of 500 μg·L-1 (pH 4, 25℃). Results: The maximum adsorption capacity of the adsorbent material was 137.3 mg·g-1 within 24 h; moreover, the uranyl removal reached 96% within 240 min. Limitations: Only simulated seawater adsorption experiments have been conducted, and real seawater adsorption experiments are yet to be conducted. Conclusions: The adsorbent had an AO utilization rate of 1/3.5 and was stable over a pH range of 4–10, with good selectivity and reusability, demonstrating its potential for seawater uranium extraction.
Subjects: Chemistry >> Nuclear Chemistry submitted time 2023-06-06
Zhang, Hang
Ao, Yin-Yong
Wang, Yue
Zhao, Shang-Jie
Sun, Jia-Yang
Zhai, Mao-Lin
Li, Jiu-Qiang
Peng, Jing
Li, Hui-Bo
Abstract： N,N,N’,N’-Tetraoctyl diglycolamide (TODGA) is one of the most promising extractants tailored for high-level liquid radioactive waste treatment during nuclear fuel reprocessing. The γ-radiolysis of TODGA (0.2 mol/L) in n-dodecane (nDD) solution with and without pre-equilibrated 3.0 mol/L HNO3 was investigated using HPLC and UPLC-QTOF-MS and compared with the γ-radiolysis of neat TODGA in this study. With increased absorbed doses, the concentration of TODGA decreased exponentially for the studied systems. Moreover, pre-equilibration with HNO3 (3.0 mol/L) slightly influenced the γ-radiolysis of TODGA in nDD. Seven radiolytic products generated from the rupture of the C – C, C – O, and C – N bonds in TODGA were identified in the studied extraction system. The influence of γ-radiation on TODGA/nDD for the extraction of Eu(III) was evaluated using the first combination of extraction experiments and density functional theory (DFT) calculations, in which the complexations of Eu(III) with TODGA and its radiolytic products were systematically compared. Based on the radiolysis kinetic model of TODGA, the slope curve of the distribution ratio of Eu(III) (DEu ) and the absorbed dose, and fluorescence titration analysis, the empirical equation of the absorbed dose and DEu was obtained successfully. Below 300 kGy, the experimental DEu agreed well with the obtained empirical equation for TODGA/nDD. Conversely, at a high absorbed dose, the experimental DEu was higher than the theoretical DEu based on the empirical equation because the radiolytic products of TODGA with similar coordination structures still possessed partial complexation toward Eu(III), which was confirmed by DFT calculations. This work provides a method to predict the extraction distribution ratio of an irradiated extractant system and to understand the complex extraction process.
Subjects: Chemistry >> Nuclear Chemistry submitted time 2023-05-31
Abstract： The rapid development of advanced techniques for selective and efficient U(VI) extraction from aqueous solutions is essential for addressing U(VI) environmental pollution and energy issues. Here, we share recent progress in U(VI) extraction from aqueous solutions, especially the most frequently applied techniques such as adsorption, catalysis (photocatalysis, piezocatalysis, and electrocatalysis), chemical deposition, and reduction by zero-valent metal particles. We attempt toelucidate the strategies and various mechanisms that contribute to the enhancement of selective U(VI) extraction. At the end of our review, we highlight the outlook, challenges, and prospects for the development of this field.
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