This study shows an implementation of neutron-gamma pulse shape discrimination (PSD) using a two-dimensional convolutional neural network. The inputs to the network are snapshots of the unprocessed, digitized signals from a BC501A detector. By exposing a BC501A detector to a Cf-252 source, neutron and gamma signals were collected to create a training dataset. The realistic datasets were created using a data-driven approach for labeling the digitized signals, having classified snapshots of neutron and gamma pulses. Our algorithm was able to successfully differentiate neutrons and gammas with similar accuracy as the Charge Integration (CI) approach. Additionally, the independent dataset accuracy for our suggested 2D CNN-based PSD approach is 99%. In contrast to the traditional charge integration method, our suggested algorithm with data augmentation, is capable of extracting features from snapshots of the raw data based on the signal structures, making it computationally more efficient and also appropriate for other types of neutron detectors.

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.

In clinical boron neutron capture therapy (BNCT), the distribution of dose to a heterogeneous medium that is predicted by a treatment planning system (TPS) should be experimentally validated. A head phantom specifically developed for this purpose is described and demonstrated herein. The cylindrical phantom exhibits distinct regions made from four materials (polymethyl methacrylate, calcium phosphate, air, and boric acid) to approximate a head structure with explicitly defined skin, skull, and brain tissue with a cavity and tumor within. Multiple gold activation wires were consecutively embedded in the phantom along the neutron beam central axis to provide direct measurements of Au(n,γ)Au reaction rates at various depths and interfaces, enabling comparison with numerical predictions. The heterogeneous head phantom was irradiated and tested at the Tsing Hua Open-pool Reactor BNCT facility, and the measured reaction rates were compared with the results of two calculation methods: Monte Carlo simulations based on an analytical geometry model and TPS predictions based on voxel-based geometry created from computed tomography images. The specifications of the phantom and reported results can serve as a reference suite of test problems for BNCT dosimetry.

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 Ra in water using Ra-Nucfilm discs. Ra 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 Ra and Ba using as internal tracer.

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.

A multi-detector soil radon measurement system based on IoT (the Internet of Things) has been developed for the specific application of long-term monitoring of soil radon concentration in remote mining areas. The system utilizes the scintillation chamber method to measure radon concentration, with SiPM (Silicon photomultiplier) for photoelectric conversion. This is combined with temperature compensation technology and 'triple-proof' protection measures to enhance the anti-interference capability of the instrument, thereby indirectly ensuring the accuracy of the measurement results. To address the issue of inconvenient data networking in the field, a complementary 'NB-IoT (Narrow Band Internet of Things) + Bluetooth' dual wireless network transmission method is employed. Additionally, the online monitoring and management platform for soil radon concentration on the cloud server enables online monitoring and management of data.The developed system demonstrated a sensitivity of 1.56 cph/(Bq/m³), a relative error of ≤10%, and an relative standard deviation (RSD) of ≤5.59%. Additionally, the system exhibited an endurance of 53 h when powered by a 12Ah battery and connected to three measurement nodes. The calibrated system has conducted long-term monitoring of radon concentration in a uranium mining area. The test and practical application demonstrate that the developed system meets the requirements of field data networking and the expansion of multiple detection nodes, operates reliably, and enables long-term continuous online monitoring of radon concentration at multiple depths of a single measuring point and multiple measuring points in a region. This provides effective data support for soil radon-related research.

Boron neutron capture therapy (BNCT) perform as a treatment option for locally advanced or recurrent unresectable head and neck cancers since June 2020 in Japan. The effect of BNCT on parotid carcinoma, which presents a variety of histologic types, remains unclear. The object of this study was to investigate the antitumor efficacy of BNCT against parotid gland carcinoma by focusing on LAT1, which is involved in the uptake of L-BPA, the boron compound used in BNCT.

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 with a heterogeneous distribution of AmO 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 AmO 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.

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), 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.

Precise measurements of fundamental decay data such as energies and transition probabilities of radioactive isotopes are important for the development of corresponding nuclear modelling, activity determination and various applications in science and technology. The EMPIR project PrimA-LTD -"Towards new Primary Activity standardisation methods based on Low-Temperature Detectors" - aims to measure the electron-capture decay of Fe very precisely using Metallic Microcalorimeters (MMCs) with outstandingly high energy resolution. Using a high-statistics measurement, electron-capture probabilities shall be precisely determined and higher-order effects such as electron shake-up and shake-off shall be examined with unprecedented precision. A key to success for this project is sample preparation. This work reports on the implantation of Fe into the 140 μm × 140 μm gold absorbers of the MMCs as a proof of principle for scalability. Building up on preparatory laser-spectroscopic studies on stable Fe, laser resonance ionization at the RISIKO mass separator was used to produce a monoisotopic Fe ion beam with the required specifications. Successful implantations of this isotope (i) into 32 test absorbers with about 0.7(2) Bq each and (ii) into various on-chip absorbers with an activity close to the requested 5 Bq per absorber are presented. The impact of the implantation on the quality of spectra is highlighted on the basis of first MMC test measurements.

In the context of upgrading the Large Hadron Collider (LHC) to its High-Luminosity (HL-LHC) configuration, it is essential to conduct a thorough zoning classification and characterization of activated cables within the particle accelerator. To address this need, a methodology was developed to identify regions where materials can be cleared from regulatory control in compliance with the Swiss Radiation Protection Legislation. The study begins with optimizing the elemental composition of cables and validating Monte Carlo FLUKA simulations using high-energy resolution gamma spectrometry (GS) and total gamma counting (TGC) measurements on 19 copper cable samples, collected during the winter shutdown 2023/2024. This methodology enables performing radiological zoning and accurately defines the radiological classification of the cables installed in the LHC Points 1 and 5, including both the accelerator tunnel and service galleries, prior to dismantling. Finally, the study proposes a conservative scaling factor for cable zoning and introduces a TGC figure of merit (FOM), representing a conservative activation scenario for the copper cable types.

This study presents the investigation of the radiation interaction properties for SS304 and Incoloy 800H alloys, which are widely used in PWRs and HTGRs. First of all, theoretical and MC simulation evaluations are performed, then experiments are conducted for further analysis. The findings indicate no significant difference in mass attenuation coefficients (MAC) and gamma-ray radiation protection efficiencies (RPE) between the two alloys. Additionally, both SS304 and Incoloy 800H exhibit similar neutron shielding capabilities, with comparable effective removal cross-sections and numbers of transmitted neutrons at different neutron energies (0.025 eV, 100 eV and 4.5 MeV). The study also examines secondary radiation generated by neutron interactions. The impact of thermal treatment (300 °C, 500 °C, 700 °C and 1000 °C) and cooling approaches (quenching and self-cooling) on these alloys were further experimentally examined. Notably, thermal treatment changes the MAC values, particularly at 1000 °C, with SS304 showing a more distinct change than Incoloy 800H. Besides, quenched samples have higher MAC values compared to self-cooled samples, especially at 1000 °C. However, the microhardness values remained largely unaffected by heat treatment, except at 1000 °C, where both alloys exhibited reduced microhardness. The study underscores that there is no significant difference in microhardness between quenching and self-cooling techniques. These results provide valuable insights for enhancing the safety and efficiency of radiation shielding materials in nuclear reactors.

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.

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.

To investigate the potential of activated carbon from palm kernel shell waste for Tc-radiolabeled nanocarbon aerosol, a new production technology for carbon-based 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 Tc radiolabelling optimization. The optimal Tc radiolabeling conditions resulted in an adsorption capacity of 21.43 ng Re/g t-PSC (8.32 GBq 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 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.

F 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 F through photon-induced channels will reduce many of the intricacies in transportation and handling. With this objective, the integral cross sections for F(γ, n)F 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.

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 H, Co Mn in the recovered [O] water irradiated with 11 MeV protons used to produce [F] fluoride by the O(p,n)F 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 Co. The highest activity concentration of about 0.1 GBq/ml has been detected in the case of tritium H. All ten samples were qualified as transitional, low-level radioactive wastes.

Boron neutron capture therapy (BNCT) is based on nuclear reactions between thermal neutron and boron-10 preferentially distributed in the cancer cells. B-boronophenylalanine (BPA) is the approved drug for treatment of oral cancers for BNCT. However, the predictive biomarkers to evaluate therapeutic efficacy and side-effects have not been clarified yet. Here we performed comprehensive analysis of mRNA expression using human oral squamous carcinoma SAS cells after BPA-BNCT. The expression of particular mRNAs including inflammatory and immune-related responses and transcription factors, namely CSF2, ATF3, MAFB, PTGS2 and TNFAIP3 was increased 24 h after neutron irradiation of therapeutic dose of BPA-BNCT. NF-κB pathway genes were also activated after BNCT. The early increase of the gene product of CSF2 gene, granulocyte-macrophage colony stimulating factor (GM-CSF), in culture supernatant of SAS cells was observed by ELISA analysis after BPA-BNCT at a setting dose of 24 Gy-eq. The GM-CSF level was also increased after equivalent dose of gamma-ray and carbon beam irradiation. GM-CSF may be involved in local and systemic early responses of BNCT for particular types of cancer.

This study presents a new mathematical model for determining the atomic number and density of materials using dual-energy gamma-ray transmission tomography. The proposed method is based on an algebraic function that relates the atomic number to the attenuation ratio, offering an innovative approach for the precise characterization of materials. The methodology employed involved the theoretical development of the model, followed by tests using data from NIST XCOM and practical experiments with a gamma-ray transmission tomograph using americium-241 (59.5 keV) and cesium-137 (662 keV) sources. Two sets of materials were used: one group of 10 elements for determining the calibration curves and another with 4 elements (graphite, magnesium, aluminum, and iron) for validating the theoretical and experimental calibration curves. Unlike existing methods, which predominantly utilize polynomial or exponential relationships, the proposed model introduces a novel algebraic approach to enhance accuracy and computational efficiency. The performance of the new model was compared with approaches used by other authors. The analyses were conducted for elements with atomic numbers between 6 and 30, covering a significant range of materials of practical and scientific interest. The results demonstrated that the proposed model presented discrepancies of less than 3.5% for the atomic number and a maximum error of 10.04% for the density, with a trend of decreasing errors as the atomic number increased.