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Elemental composition of coffee and tea commonly consumed in Saudi Arabia were assessed using neutron activation analysis. The results of conservative estimation of hazard and risk models reveal that the consumption of the examined coffee and tea doesn't pose significant elemental or radiological threats to human health. However, Due to the non-negligible hazard quotient for few elements, attention should be given to control the consumption of black tea to avoid the possible overdose risk of these elements. For more accurate evaluation of health risk, investigations on the actual extraction of elements from tea and ground coffee will be carried out in future studies. [ABSTRACT FROM AUTHOR]

Abstract Neutron tomography is gaining popularity particularly in cultural heritage research, for non-destructively analysing the inner structure of bulk metal artefacts, such as bronzes, but the induced temporary decay radiation is often considered as a drawback. However, this delayed gamma-emission can be put to good use: by performing gamma spectroscopy after neutron tomography, the interior elemental composition of artefacts can be obtained “for free”. Inspired by this, we propose a ray-tracing approach to non-invasively quantify both interior geometry and elemental composition using only a single neutron tomography experiment. This strategy aligns well with both the aim for efficient use of neutron beam time and the expectation from curators and conservators for minimal neutron irradiation. Here, we outline the core principle of this method, demonstrate the extent of its quantification capability on bulk objects of known composition by fusing neutron tomography and delayed-gamma spectroscopy data sets. We also showcase its practical application on an ancient solid-cast Indonesian bronze statuette, by which we gain insights into how the pristine inner bronze segregated into a different composition than the surrounding shell. Similarly, the method allows us to quantify the composition of a hidden offering in the statuette that consecrates the bronze for worship purposes.

The ultra-cold neutron (UCN) source being commissioned at North Carolina State University’s PULSTAR reactor is uniquely optimized for UCN production in the former graphite-filled thermal column outside of the reactor pool. The source utilizes a remote moderation design, which is particularly well suited to the PULSTAR reactor because of its high thermal and epithermal neutron leakage from the core face. This large non-equilibrium flux from the core is efficiently transported to the UCN source through the specially designed beam port in order to optimize UCN production at any given reactor power. The increased distance to the source from the core also greatly limits the heat load on the cryogenic system. A MCNP (Monte Carlo N-Particle) model of this system was developed and is in good agreement with gold foil activation measurements using a test configuration as well as with the real UCN source’s heavy water moderator. These results established a firm baseline for estimates of the cold neutron flux available for UCN production and prove that remote moderation in a thermal column port is a valuable option for future designs of cryogenic UCN sources.

The ultra-cold neutron (UCN) source being commissioned at North Carolina State University's PULSTAR reactor is uniquely optimized for UCN production in the former graphite-filled thermal column outside of the reactor pool. The source utilizes a remote moderation design, which is particularly well suited to the PULSTAR reactor because of its high thermal and epithermal neutron leakage from the core face. This large non-equilibrium flux from the core is efficiently transported to the UCN source through the specially designed beam port in order to optimize UCN production at any given reactor power. The increased distance to the source from the core also greatly limits the heat load on the cryogenic system. A MCNP (Monte Carlo N-Particle) model of this system was developed and is in good agreement with gold foil activation measurements using a test configuration as well as with the real UCN source's heavy water moderator. These results established a firm baseline for estimates of the cold neutron flux available for UCN production and prove that remote moderation in a thermal column port is a valuable option for future designs of cryogenic UCN sources. [ABSTRACT FROM AUTHOR]

Neutron tomography is gaining popularity particularly in cultural heritage research, for non-destructively analysing the inner structure of bulk metal artefacts, such as bronzes, but the induced temporary decay radiation is often considered as a drawback. However, this delayed gamma-emission can be put to good use: by performing gamma spectroscopy after neutron tomography, the interior elemental composition of artefacts can be obtained "for free". Inspired by this, we propose a ray-tracing approach to non-invasively quantify both interior geometry and elemental composition using only a single neutron tomography experiment. This strategy aligns well with both the aim for efficient use of neutron beam time and the expectation from curators and conservators for minimal neutron irradiation. Here, we outline the core principle of this method, demonstrate the extent of its quantification capability on bulk objects of known composition by fusing neutron tomography and delayed-gamma spectroscopy data sets. We also showcase its practical application on an ancient solid-cast Indonesian bronze statuette, by which we gain insights into how the pristine inner bronze segregated into a different composition than the surrounding shell. Similarly, the method allows us to quantify the composition of a hidden offering in the statuette that consecrates the bronze for worship purposes. [ABSTRACT FROM AUTHOR]

Neutron-activated 31Si is an almost pure beta emitter and is one of the short-lived radionuclides, including beta-gamma emitter 56Mn, which were created in a form of residual radioactivity in the early period after the atomic bombing of Hiroshima and Nagasaki. The features of the biological effects of internal irradiation by these radionuclides are a subject of scientific discussions and research. The publication presents data on internal radiation doses in experimental Wistar rats that were exposed to sprayed neutron-activated microparticles of 31SiO2. Doses of internal radiation could be conditionally divided into three groups according to their values. It has been found that elevated values of internal radiation doses in rats' organs/tissues as a result of exposure to sprayed 31SiO2 microparticles with initial activity of 3.2 × 107 Bq varied from 10 to 120 mGy (eyes, lungs, skin, stomach, jejunum, large intestine). The moderate dose values were in the range from 1.9 to 3.7 mGy (trachea, esophagus, ileum). The smallest doses were received by the kidney, testis, blood, cerebellum, heart, liver, cerebrum, bladder, spleen and thymus (from 0.11 to 0.94 mGy). The obtained data are important for interpreting the results of ongoing and planned biological experiments with 31SiO2 microparticles—in comparison with the previously published data on features of biological effects caused by beta-gamma emitting 56MnO2 neutron-activated microparticles. [ABSTRACT FROM AUTHOR]

The body absorbs trace elements from food, which can have positive and negative effects depending on their type and amount. The study aimed to determine the amount of trace elements found in different varieties of Iranian rice and imported varieties. The concentration of trace elements in rice samples was measured by neutron activation analysis and inductively coupled plasma (ICP-AES). Elements concentrations (mg.kg-1) were determined for aluminum (2.92-9.16), arsenic (0.064-0.156), bromine (0.24-5.20), calcium (102-981), chlorine (132-323), chromium (ND -20.4), lead (ND-0.232), cadmium (0.010-0.115), scandium (0.001-0.007), magnesium (262-519), manganese (2.97-18.50), sodium (3.99-14.30), mercury (ND-0.002), zinc (2.62-23.60). This study found that Indian rice contains higher levels of bromine, calcium, and sodium, while Pakistani rice contains higher amounts of aluminum, chlorine, lead, and mercury. Shirodi rice is known to have higher levels of arsenic and magnesium, Tarem Hashemi rice has been found to contain higher amounts of chromium, manganese, and zinc, and cadmium is found in Sadri rice. However, the amount of toxic elements in all types of rice does not pose a significant threat to human health. [ABSTRACT FROM AUTHOR]

Introduction: Neutron contamination is likely caused by the collision of high-energy photon interactions (γ, ṉ) with heavy metals used in the construction of the accelerator. This study is essential to quantify the excess dose from neutron contamination by Elekta Precis. The main object is to assess the neutron contamination with a gas dosimeter containing He-3. Material and Methods: In this study, neutron contamination was estimated at different points in and out of the treatment room. We used two types of dosimeters: a He-3 gas dosimeter (CRAMAL 31) as a neutron dosimeter and Farmer ionization chambers, and PC-electrometer (Sun Nuclear, USA) which is sensitive to photons measured photo-neutron doses. Both the neutron and Farmer dosimeters were applied in the presence and absence of acrylic plates at the same point. In this study, Monte Carlo (MC) code was utilized to prepare the correct proportion of neutron dose. Results: At different points in and out of the treatment room, neutron contamination was approximately in the range of background dose (D = 0.001 μSv). The neutron dosimeter displayed 48.792 μSv, 25.456 μSv and 28.756 μSv for 6, 10 and 15 MV photon energy, respectively. He-3 gas dosimeter showed that the neutron dose net was negligible under the treatment field. Conclusion: He-3 gas dosimeter detected more than the usual neutron dose in 6 MV photon energy than we expected. Due to the high photon flux under the radiation fields, a He-3 neutron dosimeter reported photoneutron dose. Nevertheless, the photo-neutron dose was in the range of micro Sievert (μSv). He-3 gas dosimeter was not suitable for neutron dosimetry in places with high photon fluence because of the low energy peak in the detection of neutrons. [ABSTRACT FROM AUTHOR]

Prompt-gamma activation analysis (PGAA) is an elemental analysis method based on radiative neutron capture that has a high sensitivity to chlorine (Cl). To evaluate the feasibility of replacing the conventional wet chemistry method, ASTM C1152, with PGAA, four mixtures of concrete were prepared with Cl added, ranging from a 0.004 to 0.067% mass fraction of Cl in concrete. The PGAA method detected levels of 100 µg/g Cl in concrete. While both the PGAA and C1152 methods gave results systematically below the nominal values of added Cl, the PGAA data showed excellent correlation (R² of 0.999) with the C1152 results measured on the same samples. Given that PGAA can measure Cl in concrete as well as C1152 and is faster and less labor-intensive, it can be a candidate for development as a standard method for an alternative to the latter. [ABSTRACT FROM AUTHOR]

The study used the radio-analytical technique (Neutron Activation Analysis), besides the other advanced chemical technique, which was inductively coupled plasma mass spectroscopy (ICP-MS), in order to evaluate the bioaccumulation factor (BF), of different trace elements inside the collected green macro algal samples. The results showed that BF results were based on the algal selectivity of each analyzed trace element. The study recommended the using of Ulva fasciata, as a bio-filter of nuclear cooling water system (mitigation tool), due to its selective absorption ability for certain trace elements such as; Th, Cs &Sr. [ABSTRACT FROM AUTHOR]

The elemental composition of the paint was determined by non-destructive nuclear analytical techniques, namely neutron activation analysis (NAA), charge particle activation analysis (CPAA), and energy dispersive X-ray fluorescence (EDXRF). NAA was chosen as the primary method of analysis as there is no matrix effect, requirement of a small quantity of a sample, detection capability down to ppb level, and self-validation. Ba, Ca, and Cr are detected as the major constituents, whereas Co, Hf, Mn, Y, Sr, Sb, W, Zn and Zr were detected at trace levels in the paint sample by NAA. CPAA was also utilized for the analysis, and elements like Ca, Cr, Fe, Sr, Ti and Zr were detected in the sample. The validation of the analytical results obtained in NAA and CPAA was carried out by the EDXRF technique, in which, including Si, most of the transition and high Z elements like Ca, Cr, Fe, Sr, Zr and Ba were detected. Analysis of variance was carried out to establish the statistical indistinguishability and acceptability of the analytical results obtained by these three methods. This study showcases the capability of NAA and CPAA for routine compositional analysis of paint sample without the requirement of a matrix matched standard. [ABSTRACT FROM AUTHOR]

The Joint European Torus (JET) is playing an important role in preparing for the operation of the future world's largest tokomak, ITER. In this respect, the tritium campaign (C40) and second deuterium–tritium experiment (DTE2, C41) took place in the JET during the years 2021 and 2022. In this work, a corresponding irradiation scenario was utilized for the activation calculations of eight material foils located at the JET outer long-term irradiation station (OLTIS). Neutron-induced activities and dose rates at a 30 cm distance after shutdown at specified cooling intervals were calculated with the FISPACT-II code, employing the EAF-2010 nuclear and TENDL-2021 data libraries. The Monte Carlo MCNP6.2 particle transport code equipped with the FENDL-3.1d nuclear data library was used for the calculation of the neutron flux densities. [ABSTRACT FROM AUTHOR]

Bromine-82, Potassium-42 and Copper-64 have been successfully adopted as radioactive surrogates for outdoor large area contamination training. The goal of this project was to discover new materials that could supplement potassium bromide (KBr) and copper pellets in radiological dispersal device (RDD) training events to reduce the down time of the training fields and to broaden the toolbox of the RDD surrogates at Idaho National Laboratory. Of the ten different materials investigated, sodium nitrite, gallium metal, and gallium oxide presented the greatest promise due to their ability to meet or exceed selection criteria: half-lives (between 10 and 36 h), limited impurities (99.9%), and favorable National Fire Protection Agency (NFPA) rating. [ABSTRACT FROM AUTHOR]

Even today when nearly 80 years have passed after the atomic bomb (A-bomb) was dropped, there are still debates about the exact doses received by the A-bomb survivors. While initial airborne kerma radiation (or energy spectrum of emitted radiation) can be measured with sufficient accuracy to assess the radiation dose to A-bomb survivors, it is not easy to accurately assess the neutron dose including appropriate weighting of neutron absorbed dose. Particularly, possible post-explosion exposure due to the radioactive particles generated through neutron activation have been almost neglected so far, mainly because of a large uncertainty associated to the behavior of those particles. However, it has been supposed that contribution of such non-initial radiation exposure from the neutron-induced radioactive particles could be significant, according to the findings that the stable chromosomal aberration rates which indicate average whole-body radiation doses were found to be more than 30% higher for those exposed indoors than for those outdoors even at the same initial dose estimated for the Life Span Study. In this Mini Review article, the authors explain that such apparently controversial observations can be reasonably explained by assuming a higher production rate of neutron-induced radioactive particles in the indoor environment near the hypocenter.

An instrumental neutron activation analysis method in conjunction with Compton suppression spectrometry was developed for the simultaneous determination of nanogram levels of Cl, Br and I in salmon muscle and skin. Several non-chlorinated solvent systems were evaluated for the extraction of lipids containing organohalogens and an acetone-hexane (1:1) system was selected because it gave the highest extraction efficiency with least contamination. An extensive quality assurance program was designed. Lipid fractionation procedures were developed using size exclusion, solid phase extraction and thin layer chromatography followed by characterization using 1H, 13C and 31P nuclear magnetic resonance and mass spectrometry. Results show that both muscle and skin had high levels of triacylglycerols. [ABSTRACT FROM AUTHOR]

The global effort to establish Accelerator-Based Boron Neutron Capture Therapy (AB-BNCT) facilities involves various accelerator technologies and neutron-producing targets, each characterized by different properties of the primary beam and neutron spectra they generate. With an emphasis on long-term sustainability, it is essential to minimize the production of residual radioactivity to the lowest possible level, particularly given their intended use in a hospital environment. This paper aims to quantitatively assess the residual radioactivity in these facilities, taking into account both primary and secondary activation. Primary activation primarily arises from the interaction of the proton or deuteron beam and the neutron-producing target. Secondary activation results from neutron-induced reactions on the elements exposed to the neutron flux, with the Beam Shaping Assembly (BSA) being the most exposed one. To assess activation, we evaluated a representative group of target-BSA configurations. Primary activation was calculated based on cross-sectional data and the corresponding target materials. Neutron activation was assessed using Monte Carlo simulations with the MCNP 6.1 code. Regarding target activation, our findings indicate that 9Be targets working with protons of less than 10 MeV represent the cleanest option, while 7Li targets working with protons lead to the highest activation levels. As for BSA activation, the neutron energy is a crucial factor. In the case of standard BSA materials, higher neutron energy results in an increased number of potential reactions that produce radioactive products. Additionally, our findings suggest that radioactivity induced by impurities and minor components in alloyed materials cannot be disregarded and must be taken into account in radioactivity calculations. In summary, this research provides a comprehensive analysis of activation of the commonly used targets and BSA materials, aimed at contributing to the optimization of AB-BNCT facilities from a radiological perspective.

LiF-BeF2-ZrF4-UF4 and LiF-BeF2 have been employed as fuel salt and coolant in liquid fueled molten salt reactor, respectively. A heat exchanger containing both fuel salt and cooling salt is placed inside the reactor vessel to dissipate the heat produced by the reactor. A molten salttoair heat exchanger outside the reactor vessel is used to cool the molten salt, and a molten salt pump is used to drive the circulation of the entire circuit of coolant system. As of the molten salt, LiF-BeF2 coolant, flows through the molten salt heat exchanger located inside the reactor vessel, nuclides in the coolant are activated by 19F(n, α)16N, 19F(n, γ)20F, and 19F(n, p)19O reactions with neutrons in the reactor core, and thus radionuclides such as 16N, 20F and 19O are produced. These radionuclides have radiation impact on the surrounding environment and equipment if they enter the rooms where equipment for molten salt coolant circuit locate. According to the flow process and distribution law of the cooling salt, theoretical calculation formulas to calculate the amount of radionuclide were given. Using theoretical formulas, calculated radioactivities of the three most important radionuclides 16N, 20F and 19O per unit volume of cooling salt are 9.05×106, 8.33×106 and 2.69×105 Bq/cm3 at the outlet of the molten salt heat exchanger, while the radioactivities of the three radionuclides at the same location calculated by ORIGENS program are 8.98×106, 8.58×106 and 2.76×105 Bq/cm3, respectively. The results calculated by the formulas are in good agreement with the data of the ORIGENS program, and the maximum relative deviation between the results of the two methods is 292%, indicating that applying the derived formulas to carry out cooling salt activation analysis is reasonable and feasible. These radionuclides, 16N, 20F and 19O, produced in the cooling salt of molten salt reactor release highenergy gamma rays as they decay. The total gamma emissivities of the cooling salt in equipment and pipelines located in the cooling salt storage tank room, in the molten salt-to-air heat exchanger room and in the cooling salt pump room are 8.71×1011, 9.51×1011 and 1.93×1012 s-1, respectively, and the most radioactive area is in the cooling salt pump room. The maximum absorbed dose rate at the location of the cooling salt auxiliary equipment is 45.7 mGy/h without shielding, which cannot meet the requirement of the radiation dose, and which should be less than 50 Gy for the entire life of the equipment (300 full power days). By setting a 30 cm thick concrete shielding wall between the cooling salt auxiliary equipment and the cooling salt pump, the absorbed dose rate at the location of the auxiliary equipment can be reduced to less than 7 mGy/h, thereby the radiation protection requirements for the cooling salt auxiliary equipment can be met.

Neutron sources are utilized for different aims, such as studying the material’s internal structure, investigating the materials’ crystal structures, and neutron therapy. Using more than one isotopic neutron source in some aims could be beneficial. Herein, a neutron irradiator consisting of two 241Am-9Be neutron sources has been designed. Radiation characterizations, including the neutron fluence and dose rate, were investigated. The Monte Carlo simulation code Monte Carlo N-particle transport code version 5 (MCNP5) was used to conduct all the required investigations to achieve an optimal design. Two factors were considered; the first was the dose rate due to both neutron and gamma radiations to meet radiation protection requirements; MCNP5 was used to determine the adequate thickness of the shielding material. The second parameter is the neutron fluence. Two irradiation sites have been proposed, one for fast neutron irradiation and the other for thermal neutron irradiation. By utilizing MCNP5, we have determined the radiation characterization at the two sites to be used as the irradiation sites. After the optimal design had been produced, the thermal and the epithermal neutron fluences inside the two irradiation sites were determined experimentally. The experimental results showed a perfect agreement with the simulation ones. The neutron irradiator designed from merged two isotopic neutron sources could be a benefit in the neutron radiation applications in the labs.

Characterizing neutron activation of stainless steel is a critical task for many applications, and yet little activation data is available for this alloy. To address the need, a series of neutron activation experiments have been performed using the Flattop critical assembly at the National Criticality Experiments Research Center. Several experimental conditions were probed for complete understanding of the neutron environment, including foil composition and position within the assembly. The experimental results reported here will aid in the production of stainless steel benchmarks for a range of nuclear data applications including those used in nuclear forensics analysis. [ABSTRACT FROM AUTHOR]

The Joint European Torus (JET) is playing an important role in preparing for the operation of the future world’s largest tokomak, ITER. In this respect, the tritium campaign (C40) and second deuterium–tritium experiment (DTE2, C41) took place in the JET during the years 2021 and 2022. In this work, a corresponding irradiation scenario was utilized for the activation calculations of eight material foils located at the JET outer long-term irradiation station (OLTIS). Neutron-induced activities and dose rates at a 30 cm distance after shutdown at specified cooling intervals were calculated with the FISPACT-II code, employing the EAF-2010 nuclear and TENDL-2021 data libraries. The Monte Carlo MCNP6.2 particle transport code equipped with the FENDL-3.1d nuclear data library was used for the calculation of the neutron flux densities.

In the present study, understoichiometric TiZrCN, TiZrNbCN, and TiZrSiCN coatings were produced using the cathodic arc technique with a C/N ratio of approximately 0.5 to investigate their potential use in nuclear technology. The coatings were evaluated for their corrosion resistance in 3.5 % NaCl and neutron activation. The effect of adding Si and Nb to the quaternary TiZrCN system was also investigated. The results showed that the addition of Si (∼4.64 at.%) to the matrix of TiZrCN improved their electrochemical properties in NaCl solution, the protective efficiency was 92%, while the Nb addition (∼5.5 at%) lead to the decrease in corrosion resistance by 1.39 times comparing with TiZrCN. Furthermore, after fast neutron irradiation at a nominal power of 1450 kW, none of the coatings were activated, indicating good radiation resistance. It was determined from the structural analysis that the Ti6Al4V substrate before corrosion consists of hexagonal and cubic space groups with different lattice parameters. By adding Si and Nb, a small amount of ZrO2 and Si3N4 was detected along with the main phases in the TiZrCN structure.

Radioembolization shows great potential as a treatment for intermediate- and advanced-stage liver cancer. However, the choices of radioembolic agents are currently limited, and hence the treatment is relatively costly compared to other approaches. In this study, a facile preparation method was developed to produce samarium carbonate-polymethacrylate [152Sm2(CO3)3-PMA] microspheres as neutron activatable radioembolic microspheres for hepatic radioembolization. The developed microspheres emits both therapeutic beta and diagnostic gamma radiations for post-procedural imaging. The 152Sm2(CO3)3-PMA microspheres were produced from commercially available PMA microspheres through the in situ formation of 152Sm2(CO3)3 within the pores of the PMA microspheres. Physicochemical characterization, gamma spectrometry and radionuclide retention assay were performed to evaluate the performance and stability of the developed microspheres. The mean diameter of the developed microspheres was determined as 29.30 ± 0.18 µm. The scanning electron microscopic images show that the spherical and smooth morphology of the microspheres remained after neutron activation. The 153Sm was successful incorporated into the microspheres with no elemental and radionuclide impurities produced after neutron activation, as indicated by the energy dispersive X-ray analysis and gamma spectrometry. Fourier transform infrared spectroscopy confirmed that there was no alteration to the chemical groups of the microspheres after neutron activation. After 18 h of neutron activation, the microspheres produced an activity of 4.40 ± 0.08 GBq.g−1. The retention of 153Sm on the microspheres was greatly improved to greater than 98% over 120 h when compared to conventionally radiolabeling method at ~85%. The 153Sm2(CO3)3-PMA microspheres achieved suitable physicochemical properties as theragnostic agent for hepatic radioembolization and demonstrated high radionuclide purity and 153Sm retention efficiency in human blood plasma. [ABSTRACT FROM AUTHOR]

Neutron data and cross-sections are highly regarded and are essential for developing nuclear equipment such as advanced fission and fusion reactors, accelerators, neutron shielding, physics studies, etc. The neutron cross-section should preferably be measured using a single-energy neutron beam, although the presence of a background in research reactors can affect its accurate determination. The Neutron Powder Diffraction (NPD) facility of Tehran Research Reactor (TRR) has been taken into consideration for measuring the neutron cross-section based on its properties, including neutron monochromator and multiple collimators. In this work, radiative capture cross-sections of Au, In, and Rh materials have been calculated using TRR monochromatic beam. MCNPX is a Monte Carlo particle transport code that has been applied to simulate the measurement system of the neutron cross-section and calculate the reaction rates. The effect of the presence and absence of different sections of the background on the cross-section values was investigated and the results were compared with EXFOR data library for validation. According to the findings, neutron backgrounds can have varying impacts depending on factors such as sample material, the isotope resonance regions, neutron source spatial distribution, and neutron monochromatic energy. However, the presence of fast neutron background contributes to the most uncertainty in the cross section values while its removal produces an average discrepancy from experimental libraries of 7.16%. Also, removing the cold neutron background also causes a relative difference equal to 7.65%. [ABSTRACT FROM AUTHOR]

The neutron capture cross section for the 164Dy(n,γ)165Dy reaction at 0.0372 eV energy was measured using the horizontal channel (Neutron Diffraction Facility) of Es-Salam nuclear research reactor in Birine-Algeria. The Neutron Activation Analysis (NAA) technique using 197Au(n,γ)198Au as reference reaction was adopted. The γ ray spectra for the Dy2O3 target powder and two thin gold foil were detected for the Germanium Hyper Pure coupled in the Genie 2000 data acquisition software. A resulting σ-value (2198.5 ± 94) b was measured for the thermal neutron cross section after correcting the thermal neutron self-shielding and γ attenuation effects. The measured value was then compared to the evaluated cross sections reported in two different libraries of neutron databases ENDF/B-VIII.0 and JEFF-3.3. The measured data at 0.0372 eV has been extrapolated to the energy of 0.0253 eV assuming 1/ν dependence in the thermal energy region and the result was compared with the fully measured values reported in the literature. [ABSTRACT FROM AUTHOR]

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The absorption of neutrons in media together with its transport properties cause the neutron flux to decrease as it penetrates the material because the absorption of neutrons in the sample itself attenuates the neutrons flux as it goes deeper into the sample. In the present work, the thermal neutron self-shielding factors of indium, gold, zinc, and mercury were determined experimentally. The current results together with those found in the literature were used to validate a mathematical ab initio formulae based on integral cross-section parameters used to compare our results. The complete agreement among these species of data suggests the validity of correlating the neutron migration length in the convex-shaped material with the average chord length described in the mathematical model. [ABSTRACT FROM AUTHOR]

Neutron flux parameters at pneumatic irradiation tube Pn-1 of Kyoto University Research Reactor (KUR) were measured in order to establish k0-INAA method. The measured thermal to epithermal neutron fluxes ratio (f), shape factor of epithermal neutron flux (α), and thermal (ϕth) and fast (ϕf) neutron fluxes were 29.7 ± 2.0, − 0.015 ± 0.003, and (3.17 ± 0.11) × 1012 and (4.79 ± 0.16) × 1011 cm−2 s−1 at 1 MW, respectively. The neutron flux spectrum showed a remarkable stability over the weekly operation cycle as well as at different reactor powers. The analysis results of reference materials [multi-element liquid standard, and Oriental Basma Tobacco Leaves (INCT-OBTL-5)] revealed that the k0-INAA procedure established at the KUR can be regarded as a reliable standardization method of multi-elemental INAA; measured concentrations have zeta-score < ± 2. Corrections due to nuclear interference resulting from fast neutron reactions must be considered for samples with relatively high concentrations of interfering elements. MCNP5 code was used to estimate the total efficiencies of gamma-rays, which might be required to calculate correction factor for the true coincidence effects. [ABSTRACT FROM AUTHOR]

Determination of impurities in graphite is very important for its quality control, as their presence even at trace level can affect the performance of graphite in various applications. Graphite with equivalent boron content (EBC) less than 5 mg kg-1 is considered as nuclear grade. Elements with high neutron absorption cross section (boron and rare earths) contribute significantly to EBC. Non-destructive method is preferred as there is no sample processing and probability of loss of volatile elements while digestion. Proton Induced Gamma Ray Emission (PIGE), Instrumental Neutron Activation Analysis (INAA) were utilized for the non-destructive determination of impurities in both nuclear and commercial grade graphite. Low Z elements like Li, B, F, Na, Al and Si were detected in graphite by PIGE whereas Na, K, Sc, Cr, Mn, Fe, Co, Zn, Rb, Zr, Sb, Cs, La, Ce, Nd, Sm, Eu, Tb, Yb, Hf, Ta, Th were determined using INAA. Few elements like Ca, Ti, V, Ni, Sr and Pb remained undetected by both the non-destructive techniques. These elements were determined by Total Reflection X-ray Fluorescence (TXRF) after digestion of the graphite samples by dry ashing. Combinations of these techniques were utilized to get maximum information regarding the impurities present in graphite. [ABSTRACT FROM AUTHOR]

Spallation targets for neutron production with high energy protons are made of high density and high atomic number materials in order to maximise the yield of neutrons for all the instruments around. Operating a proton beam onto a spallation target produces residual radioactive nuclei either as direct product of the spallation process and as secondary low energy neutron absorption. A reliable estimation of the overall activation and decay heat, as a function of the cooling time and irradiation profile history, is fundamental for a valuable design of the radiation shielding and cooling system during the operation phase as well for envisaging the optimal storage solution at the end of life of the target. This work presents the comparison between the FLUKA predictions of the decay heat in the ISIS TS1 target operated between 2014 and 2019 and the decay heat estimations derived from the measurement of the temperature in each plate at different cooling times. The agreement between the FLUKA predictions and the experimentally assessed values shows and quantifies the goodness of the FLUKA model in predicting measurable physical quantities relevant for the engineering thermal design of the target/reflector and moderator (TRAM) assembly. In addition, it also provides an indirect evidence of the accuracy of the simulated spallation physics and neutron transport throughout the TRAM assembly. Finally this work attempts to highlight and propose a general empirical procedure that could be eventually applied and used to proficiently measure the decay heat at whatever cooling time in targets with similar ISIS design. [ABSTRACT FROM AUTHOR]

To enhance the biological effects of radiation damage in cancerous cells, we present an alternative approach to the use of gold nanoparticles (AuNP), focusing on the synthesis and characterization of highly monodisperse, spherical radioactive gold nanoparticles 198AuNP. The size of the AuNP size was optimized with the help of Geant4/TOPAS particle scattering simulations, and energy deposition per nm3 per decay for varying radii (2–10 nm) was evaluated. This work is the foundation for ongoing experimental work to evaluate cell death induced by 198AuNP which aims for the use of radioactive gold nanoparticles in cancer treatment. [ABSTRACT FROM AUTHOR]

We present NEAR, a new experimental area at the CERN-n_TOF facility and a possible setup for cross section measurements of interest to nuclear astrophysics. This was recently realized with the aim of performing spectral-averaged neutron-capture cross section measurements by means of the activation technique. The recently commissioned NEAR station at n_TOF is now ready for the physics program, which includes a preliminary benchmark of the proposed idea. Based on the results obtained by dedicated Monte Carlo simulations and calculation, a suitable filtering of the neutron beam is expected to enable measurements of Maxwellian Averaged Cross Section (MACS) at different temperatures. To validate the feasibility of these studies we plan to start the measurement campaign by irradiating several isotopes whose MACS at different temperatures have recently been or are planned to be determined with high accuracy at n_TOF, as a function of energy in the two time-of-flight measurement stations. For instance, the physical cases of 88Sr(n, γ), 89Y(n, γ), 94Zr(n, γ) and 64Ni(n, γ) are discussed. As the neutron capture on 89Y produces a pure β -decay emitter, we plan to test the possibility to perform activation measurements on such class of isotopes as well. The expected results of these measurements would open the way to challenging measurements of MACS by the activation technique at n_TOF, for rare and/or exotic isotopes of interest for nuclear astrophysics. [ABSTRACT FROM AUTHOR]

Neutron-activated 31Si is an almost pure beta emitter and is one of the short-lived radionuclides, including beta-gamma emitter 56Mn, which were created in a form of residual radioactivity in the early period after the atomic bombing of Hiroshima and Nagasaki. The features of the biological effects of internal irradiation by these radionuclides are a subject of scientific discussions and research. The publication presents data on internal radiation doses in experimental Wistar rats that were exposed to sprayed neutron-activated microparticles of 31SiO2. Doses of internal radiation could be conditionally divided into three groups according to their values. It has been found that elevated values of internal radiation doses in rats' organs/tissues as a result of exposure to sprayed 31SiO2 microparticles with initial activity of 3.2 × 107 Bq varied from 10 to 120 mGy (eyes, lungs, skin, stomach, jejunum, large intestine). The moderate dose values were in the range from 1.9 to 3.7 mGy (trachea, esophagus, ileum). The smallest doses were received by the kidney, testis, blood, cerebellum, heart, liver, cerebrum, bladder, spleen and thymus (from 0.11 to 0.94 mGy). The obtained data are important for interpreting the results of ongoing and planned biological experiments with 31SiO2 microparticles-in comparison with the previously published data on features of biological effects caused by beta-gamma emitting 56MnO2 neutron-activated microparticles., (© The Author(s) 2024. Published by Oxford University Press on behalf of The Japanese Radiation Research Society and Japanese Society for Radiation Oncology.)

The Juru rivers are highly industrial, urbanized and agricultural. The aim of the study was to examine the geochemical bases and enrichment of rare earth elements in Juru mangrove sediments. This study is carried out to investigate the concentrations of rare earth element (REEs) elements Samarium (Sm), Europium (Eu), Yttrium (Yb), Lutetium (Lu), and Terbium (Tb) present in surface mangrove sediments of Juru River, West Coast of Peninsular Malaysia. The concentrations of rare earth elements in the sediments were determined by using the instrumental neutron activation analysis (INAA) method. For calibration and quality control procedures, standard reference material SL-1 and blank samples were irradiated together with sediment samples. The degree of anthropogenic impact on sediments were computed using enrichment factor. The concentration and enrichment factor of all the REE elements ranged between 0.42 mg/kg (Lu), 7.55 mg/kg (Sm), 0.77 (Lu) and 2.96 (Tb), respectively. The geochemical behavior of REEs in surface sediments and normalized patterns (chondrite and shale) have been studied. These findings indicate that the level of pollution has not reached an extreme or severe level, but ongoing studies should be carried out on the inputs of anthropogenic activities at Juru river. [ABSTRACT FROM AUTHOR]

The cross sections of the reactions 70Zn(n,2n)69mZn,74Ge(n,α)71mZn and 90Zr(n,2n)89m,gZr in the energy range of 13.82–14.71 MeV were measured by the activation technique in combination with HPGe detector γ-ray spectroscopy. The measured results were compared with other experimental data and with the data given in the library TENDL-2019. A nuclear model calculation based on the code TALYS-1.8 was also performed after adjustment of an input parameter for the spin distribution of level density. The results of this work strengthen the database and could be useful in further evaluation of the data. [ABSTRACT FROM AUTHOR]

In this study, we used the instrumental neutron activation analysis technique (INAA) for analysis of granite samples from four locations in Oulmes region of central Morocco. In order to investigate the geochemical behavior of both major and trace elements in granite rocks from study area. The four samples almost accurately reflect the geochemical variety and heterogeneity of the formations were taken from distinct outcrops. The aim of study was to characterize the mineralization patterns of granitic rocks, their sources and basic composition. The results highlight the importance of nuclear techniques in geochemical investigations and provide some insights into the geochemical processes that shaped these rocks.

A challenging issue for the magnetic fusion concept of generating nuclear power is the performance of plasma-facing components, particularly in the divertor within the fusion power plant. Tungsten alloys, such as W-TiC, W-La2O3, etc., are the prime candidate structural materials for advanced He-cooled divertors in next-step devices beyond ITER. In recent years, W-Ni-Fe heavy alloys (WHAs) have emerged as more promising alternatives than the above-mentioned alloys. WHAs exhibit superior fracture toughness and better fabricability. However, Ni raises concerns because it may generate an unsafe level of radioactive waste (radwaste). Given the significant promise of WHAs, neutron activation assessment is critically needed, particularly to determine the limit of Ni that can be employed. This is the main purpose of this study. The assessment is provided for disposal option and for recycling alternative of divertors made of WHAs. Additional calculations are performed to obtain the neutron flux and spectrum, transmutation products, and He/H gas production data for a typical divertor design. These additional data are generated to inform future multiscale simulations of microstructure evolution and radiation damage accumulation in WHAs. Furthermore, estimation of atomic displacement damage dose is presented. In the future, the estimated dose should be compared with the more accurate data from the multiscale simulations and revised accordingly.

The knowledge of radioactive 14C yield under atmospheric thunderstorm flash conditions (the additional channel of 14C production relative to the main cosmogenic one) is important for radiocarbon analysis. A gross model for evaluation of the thunderstorm 14C yield simulated for the altitudes up to 15 km is proposed. It was observed that yield from the thunderstorm mechanisms of 14C creation cannot compete with cosmogenic production which is six orders of value larger. The obtained result allows us to eliminate the problematic issue on thunderstorm radiocarbon generation in the atmosphere as the additional significant source. Plain Language Summary: The creation of isotopes takes place not only in stellar conditions but also in the Earth's atmosphere under cosmic irradiation. Radioactive carbon 14C produced in the atmosphere is an exclusively important tool for historical dating (in archeology, glaciology, biology, paleontology, geology, Sun activity and climate in the past) for the time scale up to ∼(50–60) thousands of years. But radiocarbon creation is also possible from thunderstorm discharges. In case of significant yield, it will cause the correction of dating results. To solve this problem, the model that takes into account the rate of 14C production depending on the part of energetic electrons in the thunderstorm flash discharges for different altitudes is proposed. The probability of 14C creation depending on the energy of neutrons generated under thunderstorms was calculated. The results revealed that yield of thunderstorm mechanism is very small when compared to the main one originated from the cosmic irradiation. It allows us to eliminate the problematic question on the correction arising from this additional source of 14C isotope. The reliability of the obtained results is confirmed by simulation of the Japan experiment on neutron registration at the strong thunderstorm in January 5 2012. Key Points: The gross model for evaluation of the radiocarbon 14C creation during thunderstormsProbability of 14C generation in the atmosphere depending on the neutron energyRadiocarbon creation under thunderstorms at different altitudes. Comparison with 14C yield from the cosmic irradiation [ABSTRACT FROM AUTHOR]