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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.

A procedure for measuring the rhenium content in radioactive ore rocks is tested. Samples of rhenium-containing rocks are irradiated with neutrons from a photoneutron source. Analysis of activation gamma spectra shows that rhenium can be identified against a background of accompanying elements, making neutron activation analysis an effective way of determining rhenium content.

Certain proton reactions emit fast neutrons, and this study examines the feasibility of measuring the neutrons emitted from a typical medical 18 MeV proton cyclotron using prompt gamma neutron activation on chlorine. This neutron detection method, employed by previous studies, has been deemed successful and highly promising. The measurement system includes a gamma radiation scintillator, for example NaI(Tl), with a spectrometry setup facing a convertor that contains a chlorine solution or compounds thereof. In its natural state, elemental chlorine is a chemically unstable gas, making its use as a convertor challenging. Therefore, chlorine is usually dissolved in solutions. Such solutions are diluted with water and contain hydrogen atoms that interact intensively with the neutrons and may interrupt their detection. Accordingly, in this study, a 25.5 L container of KCl powder was placed in front of a NaI(Tl) (7.62 cm diameter by 7.62 cm length) spectrometer near a cyclotron facility. The gamma radiation spectroscopy allowed the identification of the chlorine INS gamma line at 788 keV, alongside the K-40 gamma peak. In addition, it was possible to detect the fast neutrons using a total-counting mode, when the INS counts appear above the K-40 counting level. The chlorine INS peak in the KCl convertor was detected simultaneously during the cyclotron operations and provided a direct indication of neutron exposure around the cyclotron. We conclude that using KCl as a converter, with a NaI(Tl) gamma rays counter, is successful in detecting cyclotron-produced neutron emissions.

This study developed a manufacturing process that uses the doctor blade method for printing of neutron activation foil arrays. The doctor blade method addresses some of the limitations of other thick-film electrode printing techniques, including printing failure. Printed activation arrays of 9,216 and 36,864 foils arranged within a 10 cm × 10 cm area were prepared using the doctor blade method to improve the measurement accuracy of the neutron spatial distribution. Mass values were measured within 1% variation, indicating uniformity of mass distribution. No printing failure other than simple defects was observed. Moreover, this relatively simple process offers excellent mass-production potential.

Heavyweight concrete is mostly used for its shielding properties in the nuclear power plants. These properties can already be influenced by the selection of the input materials. In the present study, concrete samples comprised of four-component binders based on CEM I 42.5 R, blast furnace slag, metakaolin and limestone and a mixture of barite and magnetite aggregate, were investigated. Based on Energy Dispersive X-ray Fluorescence, Neutron Activation, and Prompt-Gamma Activation analyses, three concrete designs were prepared and tested. Mechanical, physical (namely cubic compressive strength, bulk density, longitudinal deformation, and dynamic modulus of elasticity) and thermal properties (thermal conductivity coefficient, specific heat capacity, and thermal diffusivity), which should be influenced by the long-term exposure to irradiation were investigated. Presented results confirmed that the prepared samples are heavyweight concrete with bulk density higher than 3400 kg.m-3 with a low level of longitudinal deformation (between 0.265 ‰ and 0.352 ‰). All the prepared samples belong to the C 35/45 concrete strength class.

In this study, 12 trace elements were measured in the shell and muscles of two species of marine and farmed Jinga shrimp Metapenaeus ensis and Litopenaeus vannamei in the Persian Gulf zone. For this purpose, the instrumental neutron activation analysis and gamma- ray spectrometry were used. The Isfahan Miniature Neutron Source Reactor as the source of thermal neutrons and high purity germanium detector for gamma-ray spectrometry was used. The aim of this study was to determine useful and unnecessary elements in farmed and marine shrimps in Persian Gulf region. The results show that the concentrations of aluminum, calcium, magnesium, manganese, bromine, and scandium in the shell samples are significantly higher than in the muscles samples. Also, shrimp muscles are a good source of useful elements such as zinc, magnesium, and manganese. And the amount of arsenic and aluminum in it is negligible.

Fusion neutron activation experiments have been carried out to investigate the activation of metal materials in Experimental Advanced Superconducting Tokamak (EAST). Some cylindrical samples made of stainless steel 316L, carbon steel (A3), high purity cadmium (Cd), oxygen-free copper (Cu), and high purity aluminum (Al) were transported to Port G in the EAST hall for neutron irradiation. The characteristic gamma rays emitted by activated samples were measured with a spectrometer consisting of a digital multichannel analyzer (MCA) and a high purity germanium (HPGe) detector. Qualitative identification of the dominant radionuclides and quantitative measurement of the specific activity have been performed. The experimental results indicated that the high radioactive nuclides with short half-life were produced, such as 56Mn in 316L and A3, 28Al in Al, 66Cu in Cu, and 111m Cd in Cd. Moreover, the long half-life radionuclides were generated in 316L, A3, Cu and Cd samples, which will become the major factors on environment and personnel. In addition, the normalized dose rate at 10 cm in the air of 316L, A3, Al, Cd, and Cu was $1.53\times 10^{-23}$ , $4.83\times 10^{-24}$ , $5.57\times 10^{-23}$ , $4.90\times 10^{-\mathrm {24}}$ and $1.10\times 10^{-\mathrm {23}}$ Sv/h/n, respectively. Finally, the comparison between the experimental neutron activation results and the predicted quantities calculated with FISPACT-2007 was discussed.

Due to the harsh radiation, thermal, and electromagnetic environment of the irradiation end of the ITER neutron activation system, which also forms a primary vacuum boundary and a nuclear safety confinement boundary simultaneously, installing a capsule detection sensor at this location, which is essential for reliable operation of the system is extremely challenging. Alternative capsule detection methods utilizing sensors located in more benign environments are investigated experimentally in this study. The experimental result suggests that a simultaneous measurement of the pressure and the acoustic/vibration wave will provide reliable information on the capsule arrival at the irradiation end.

Manufactured nanoparticles, such as CeO2, give rise to novel risks when released into the environment. To assess these risks it is important to quantify the nanoparticle mass flows, as well as their speciation and the mechanisms of their transformation. We developed an innovative dual-radiolabelling strategy for CeO2 nanoparticles using neutron activation and in-diffusion labelling to radiolabel CeO2 nanoparticles with both Ce-141 and Ce-139. The different distribution of the radiolabels in the particles does not only allow easy dose determination in uptake studies but also enables us to track the uptake pathways of the anthropogenic cerium. By measuring the activity as well as the isotope ratio we tracked the uptake, transformation and excretion of CeO2 nanoparticles in freshwater shrimp. We found that 99.99% of the uptaken particles are excreted, leaving the gut with excrement. The remaining 0.01% were internalized via a dissolution-based pathway and accumulated in the hepatopancreas of the shrimp at a dose range of pg CeO2 per shrimp. Most importantly, our results show that dissolution is not only coincidental but instrumental in the uptake of the cerium into the internal organs of the shrimp.

Concrete is a common structural material used in nuclear reactors for biological shielding, containment buildings, and waste storage ponds. Calcium is a significant component of concrete, with neutron activation of stable 40Ca (96.94% abundance) leading to the production of 41Ca (half-life 1.002(17) × 105 years). The long half-life combined with the significant amount of concrete at nuclear facilities makes 41Ca a key radionuclide for characterisation of low and intermediate level waste, as well as for long-term waste monitoring. Measurement of 41Ca by inductively coupled plasma mass spectrometry (ICP-MS) is theoretically achievable, but has previously been prevented by multiple interferences, some of which cannot be removed by offline chemical separation. This study tests the potential of tandem ICP-MS/MS and its enhanced interference separation for the detection of 41Ca in concrete samples. The instrument setup was initially optimised using stable element standards, with a combination of NH3, H2 and He proving to be most effective, achieving an instrument detection limit of 0.32 Bq g−1 (0.099 ng g−1). The procedure was validated on several concrete samples, which required the optimisation of automated borate fusion sample digestion and four-stage chemical separation, with recoveries of 73–77%. Successful detection of 41Ca was achieved at activity concentrations down to 0.67 Bq g−1 (0.21 ng g−1) at 41Ca/40Ca ratios of 10−6–10−7. This represents the first known measurement of 41Ca using ICP-MS, and recommendations are made on how instrument capabilities can be further improved so that routine measurement can be realised.

Background: Reactor pressure vessel (RV) with internals (RVI) are activated structures by neutron irradiation and volume contaminated wastes. Thus, to develop safe and optimized disposal plan for them at a disposal site, it is important to perform exact activation calculation and evaluate the dose rate on the surface of casks which contain cut-off pieces. Materials and Methods: RV and RVI are subjected to neutron activation calculation via Monte Carlo methodology with MCNP6 and ORIGEN-S program—neutron flux, isotopic specific activity, and gamma spectrum calculation on each component of RV and RVI, and dose rate evaluation with MCNP6. Results and Discussion: Through neutron activation analysis, dose rate is evaluated for the casks containing cut-off pieces produced from decommissioned RV and RVI. For RV cut-off ones, the highest value of dose rate on the surface of cask is 6.97 × 10-1 mSv/hr and 2 m from it is 3.03 × 10-2 mSv/hr. For RVI cut-off ones, on the surface of it is 0.166 × 10-1 mSv/hr and 2 m from it is 1.04 × 10-1 mSv/hr. Dose rates for various RV and RVI cut-off pieces distributed lower than the limit except the one of 2 m from the cask surface of RVI. It needs to adjust contents in cask which carries highly radioactive components in order to decrease thickness of cask. Conclusion: Two types of casks are considered in this paper: box type for very-low-level waste (VLLW) as well as low-level waste (LLW) and cylinder type for intermediate-level waste (ILW). The results will contribute to the development of optimal loading plans for RV and RVI cut-off pieces during the decommissioning of nuclear power plant that can be used to prepare radioactive waste disposal plans for the different types of wastes—ILW, LLW, and VLLW.

Nuclear reaction rates in the nickel and uranium detectors placed at different positions within the reactor cores of the TIRAN nuclear reactor complex are measured. The measured reaction rates are found to be consistent with those simulated using the PRISMA-D Monte Carlo code with the values of neutron constants as quoted in the ENDF-BVI library. Relying on the simulation, the signals of nickel neutron activation monitors are more precisely related to energy releases in the reactor core.

The results of precision experiments concerning neutron passage through the layers of lithium-7 hydride are reported. Activation integrals for neutron activation detectors positioned at various points of the composite lithium-7 hydride reflector mounted at the end of the uranium breeder system have been determined. Spectral indices have been calculated on the basis of the data obtained.

An online computer program for an interactive database was specifically developed for neutron activation analysis (NAA). The software written in Java includes gamma-ray catalog, neutron activation calculator for activity calculations and decay schemes. For the novice, expert and educator this program can greatly reduce needed information in NAA.

The thermal-neutron capture cross sections (σ0) and resonance integrals (I0) were measured for the 243Am(n,γ)244gAm and 243Am(n,γ)244m+gAm reactions by a neutron activation method. Americium-243 sa...

Rock aggregates are major ingredient of concrete used for shielding against ionising radiation generated during the operation of nuclear particle accelerators. Their analysis for trace elements which may lead to neutron activation products during prolonged operation becomes important. Proper selection of ingredients helps in minimizing the radioactive concrete waste requiring disposal at the time of decommissioning.This work involves the collection of 42 samples of seven broad types of rock aggregates and their assessment for natural radioactivity (226Ra, 232Th and 40K), lithium and other trace elements that may lead to the formation of gamma emitting, long lived neutron activation products.

The 232Th(n,γ)233Th neutron capture reaction cross-sections at the neutron energies of 4.36 ± 0.06 MeV, 3.52 ± 0.41 MeV and 2.06 ± 0.34 MeV have been measured using the activation method and off-line gamma-ray spectrometric technique. The neutrons were generated by T(p,n)3He reaction and the 197Au(n, γ)198Au was used as a monitor reaction to determine neutron flux. The effects of multiple scattering, flux attenuation, gamma-ray self-absorption and neutron flux fluctuation have been corrected in the experiment. The excitation function of the 232Th(n,γ)233Th reaction was also theoretically calculated using the TALYS-1.9 code. The experimental results were compared with the evaluated data, the calculated data and the previous experimental data.

We report here the results of a measurement of the scattered versus unscattered neutron fluence on polyethylene determined via neutron activation of multiple natural indium foils from a deuterium-deuterium (D-D) neutron generator. The neutrons were produced by the High Flux Neutron Generator (HFNG) at the University of California, Berkeley, a specially designed source to maximize neutron flux on a sample while minimizing the total neutron yield. During the experiment, approximately 108 n/s were produced with the energies at the indium foils ranging from 2.2 to 2.8MeV. Both the angle-integrated and the partial angle differential results are consistent with the predictions of the Monte Carlo N-Particle Transport (MCNP) code, using ENDF/B-VII.1. This supports shielding calculations in the fast energy region with high-density polyethylene.

Cement is an important component of concrete used as a shielding material in nuclear accelerators and reactors. Hence cement samples should be analysed for the presence of certain trace elements that may get activated by neutrons emitted during the production of radioisotopes in an accelerator, so as to minimize the low level radioactive waste to be handled during decommissioning. With this motivation the present work was undertaken and 44 samples of five broad classes of cements were analysed for natural radioactivity (226Ra, 232Th & 40K) and trace elements capable of generating long lived gamma radioactivity due to neutron activation.

The fission reactions of 238U induced by D-T neutrons were investigated by neutron activation technique and low-background HPGe-γ spectrometer. The half-life was used to identify the possible daughter nuclide. The fission cross-sections of 238U(n,f)89Rb reaction induced at the neutron energy around 14 MeV, i.e., 14.1 ± 0.3, 14.5 ± 0.3 and 14.7 ± 0.3 MeV, were measured precisely. The neutron flux was monitored by accompanying α-particle of the T(d,n)4He reaction and the neutron energies were determined by the cross-section ratio of 90Zr(n,2n)89Zr to 93Nb(n,2n)92mNb reaction. The fission cross-sections of the 238U(n,f)89Rb reaction are 33.2 ± 2.1 mb, 32.2 ± 2.0 mb, and 32.6 ± 2.1 mb, corresponding to the above neutron energy points.

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.

Materials selection for rare event physics requires high performance detectors and customized analyses. In this context a novel $\beta$-$\gamma$ detection system, comprised of a liquid scintillator in coincidence with a HPGe, was developed with the main purpose of studying ultra trace contamination of uranium, thorium and potassium in liquid samples. In the search for $^{238}\textrm{U}$ contaminations through neutron activation analysis, since the activation product $^{239}\textrm{Np}$ decays to a relatively long-lived isomeric state of $^{239}\textrm{Pu}$, it is possible to perform a time selection of these events obtaining a strong background suppression. Investigating the time distribution of the coincidences between the $\beta^-$ decay of $^{239}\textrm{Np}$ and the delayed events following the de-excitation of the $^{239}\textrm{Pu}$ isomeric level at 391.6 keV, a precision measurement of the half-life of this level was conducted. The half-life of the 391.6 keV $^{239}\textrm{Pu}$ level resulted $190.2 \pm 0.2$ ns, thus increasing the precision by about a factor of 20 over previous measurements., Comment: Accepted for publication in PRC

The current work is devoted to developing a system for the complex research of metal–hydrogen systems, including in an in situ mode. The system consists of a controlled gas reactor with a unique reaction chamber, a radioisotope positron source, and a positron annihilation spectroscopy complex. The use of the system enables in situ investigation of the defect structure of solids in hydrogen sorption–desorption processes at temperatures up to 900 °C and pressures up to 50 bar. Experimental investigations of magnesium and magnesium hydride during thermal annealing were carried out to approve the possibilities of the developed complex. It was shown that one cycle of magnesium hydrogenation–dehydrogenation resulted in the accumulation of irreversible hydrogen-induced defects. The defect structure investigation of the magnesium–hydrogen system by positron annihilation techniques was supplemented with a comprehensive study by scanning electron microscopy, X-ray diffraction analysis, and hydrogen sorption–desorption studies.

Increased reactivity of radiation-damaged quartz in the alkaline pore water of concrete and its role in radionuclide transport is a subject that is becoming increasingly important as nuclear power plants (NPPs) around the world reach end-of-life. Quartz contributes more than half the weight of most concretes and up to 40 % of the total radioactive waste volume associated with the dismantling of an NPP. The gamma-emitting radionuclides are the major contributors to the radiation dose rate in the concrete biological shield in the intermediate term after NPP shut-down. The appreciable activity and associated half-lives of these radionuclides pose a problem not only during the safe dismantling of NPPs, but also in terms of processing the concrete to reduce the active waste volume for final disposal in a repository. While thermal neutrons are captured by atomic nuclei according to their thermal neutron cross-sections, fast neutrons undergo collision cascades, which cause defects in the structure of materials. The concentration of defects accumulates as a function of neutron energy and fluence. The high covalency and the density of Si-O bonds in tectosilicates, primarily quartz, followed by feldspars, makes these minerals very susceptible to radiation damage. In this study, two inactive concrete powder samples were placed inside an instrumentation channel passing through the first biological shielding of an operating German NPP, respectively for a half- and a full fuel cycle. Using a plant-specific 3D-reactor model in conjunction with Monte Carlo neutron transport calculations, neutron fluences at the sample positions and the resulting gamma activities are calculated. We compare the measured and calculated activities of the gamma-emitting radionuclides Synchrotron X-ray powder diffractometry was used to examine structural changes on sub-nanoscale. The local structure of the silicate aggregate minerals was investigated by solid-state magic angle spinning nuclear magnetic resonance (NMR) Spectroscopy using 27Al and 29Si as probes. With this work we wish to gain insights into the potential for alkali-silica reaction (ASR) in radiation-damaged concrete and the role this plays in the transport or sorption of radionuclides either present in situ via neutron activation or ex situ due to transport of leaked reactor cooling water.

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.

This work aims to investigate the 64Cu isotope applicability for positron annihilation experiments in in situ mode. We determined appropriate characteristics of this isotope for defect studies and implemented them under aggressive conditions (i.e., elevated temperature, hydrogen environment) in situ to determine the sensitivity of this approach to thermal vacancies and hydrogen-induced defects investigation. Titanium samples were used as test materials. The source was obtained by the activation of copper foil in the thermal neutron flux of a research nuclear reactor. Main spectrometric characteristics (e.g., the total number of counts, fraction of good signals, peak-to-noise ratio) of this source, as well as line-shaped parameters of the Doppler broadening spectrum (DBS), were studied experimentally. These characteristics for 64Cu (in contrast to positron sources with longer half-life) were shown to vary strongly with time, owing to the rapidly changing activity. These changes are predictable and should be considered in the analysis of experimental data to reveal information about the defect structure. The investigation of samples with a controlled density of defects revealed the suitability of 64Cu positron source with an activity of 2–40 MBq for defects studies by DBS. However, greater isotope activity could also be applied. The results of testing this source at high temperatures and in hydrogen atmosphere showed its suitability to thermal vacancies and hydrogen-induced defects studies in situ. The greatest changes in the defect structure of titanium alloy during high-temperature hydrogen saturation occurred at the cooling stage, when the formation of hydrides began, and were associated with an increase in the dislocation density.

The composition and content of the elements of the aerial parts of the Lotus corniculatus L. wild-growing and cultivated in the Tomsk region was studied. The elemental composition was studied by neutron activation analysis and flame photometry. The neutron activation method was studied on a CANBERRA analyzer system with a pure germanium detector of the GX-3518 type with Co60 (1332 keV) 1.8 keV placement (neutron flux density 2·1013 neutrons/(cm2·s) for 12 h) . The sodium and potassium contents were determined in the emission mode by flame photometry on a Thermo Solaar M5 Mkll double-beam atomic absorption spectrometer. In wild-growing and cultivated species, the content of 31 elements (Sm, Ce, Ca, Lu, U, Th, Cr, Yb, Au, Hf, Ba, Sr, Nd, As, Ag, Br, Cs, Tb, Sc, Rb, Fe, Zn, Ta, Co, Na, K, Eu, Sb, La, Pb, Cd). In wild-growing and cultivated samples from the Tomsk region, high concentrations of Ca, K, Na, and Zn are noted, compared with the data given in the literature on lamb from other regions. According to the content of heavy metals, wild-growing and cultivated in the Tomsk Oblast baby candy meets the requirements of regulatory documents and can be used for the production of dietary supplements, herbal medicines and herbal pharmaceutical substances. In terms of Zn content, wild-growing and cultivated lamb does not meet the requirements for the minimum allowable levels of this element for feed, which manufacturers should take into account.

Data on concentrations and distribution of tritium in soils of the “Atomic” Lake excavation explosion area of Semipalatinsk Nuclear Test Site (Kazakhstan, East Kazakhstan oblast) are discussed. The concentrations of tritium in surface soil horizons and in soil profiles have been determined in the bulk soil samples and in different particle-size fractions. Tritium concentration in soils reaches 240 000 Bq/kg. A correlation has been found between tritium and europium concentrations in soils. Presumably, tritium formation is controlled by two mechanisms. One of them is the neutron activation process, i.e., nuclear reactions with lithium, boron, and nitrogen isotopes present in the soil; the other mechanism is related to the capture of tritium that was initially in the charge of mineral particles formed upon condensation from the red-hot explosion zone.

An experiment in which the fission of uranium-238 upon bombardment with neutrons is demonstrated in the context of an undergraduate lab course. The occurrence of fission is shown by means of gamma spectroscopy which reveals the presence of the fission product strontium-92. In addition, we also investigate the reaction in which plutonium-239 is formed after neutron activation of the uranium, with neptunium-239 as an intermediary step. From the gamma spectra, it is found that both reactions are demonstrated with a significance of more than 5 σ. For the fission reaction, a cross section of ( 2.8 ± 1.4 ) × 10 − 2 barn is inferred.An experiment in which the fission of uranium-238 upon bombardment with neutrons is demonstrated in the context of an undergraduate lab course. The occurrence of fission is shown by means of gamma spectroscopy which reveals the presence of the fission product strontium-92. In addition, we also investigate the reaction in which plutonium-239 is formed after neutron activation of the uranium, with neptunium-239 as an intermediary step. From the gamma spectra, it is found that both reactions are demonstrated with a significance of more than 5 σ. For the fission reaction, a cross section of ( 2.8 ± 1.4 ) × 10 − 2 barn is inferred.

The neutron capture cross-sections of 238U at the neutron energies of 4.38 ± 0.05 MeV, 3.02 ± 0.49 MeV and 2.04 ± 0.26 MeV have been measured using the activation method and off-line gamma-ray spectrometric technique. The effects of neutron flux fluctuation, multiple scattering, flux self-shielding and gamma-ray self-absorption were corrected. The excitation function of the 238U(n,γ)239U reaction was also calculated using the TALYS-1.9 code. The experimental results were compared with the evaluated data, the theoretical data and the previous experimental data.

The neutron activation properties of the PE-B4C-concrete recently developed for the European Spallation Source (ESS) ERIC (European Spallation Source, https://europeanspallationsource.se/about) were investigated. On the one hand the concrete activation was compared to that of the ordinary concrete from which it was developed by means of irradiating concrete samples in the Budapest Research Reactor (BRR) (Budapest Neutron Centre, https://www.bnc.hu/). On the other hand, the measured activities were used to study the impact of input composition on Monte Carlo activation simulations. For this purpose, the complete course of the irradiation experiments were reproduced with MCNPX and Cinder90 simulations with nominal and measured elemental concrete compositions. Simulations suggest that for realistic activation predictions more detailed elemental compositions are required than the nominal ones. Energy-dispersive X-ray fluorescence (EDXRF) analysis technique was applied for this purpose, providing fair results for short-term activation estimations.

A modification of the neutron activation method for the diagnostics of deuterium–tritium plasma of powerful pulsed fusion facilities is proposed. The method allows one to determine the effective plasma ion temperature by width of the spectral distribution of generated DT neutrons. The physical basis of the method and conditions for its implementation are described and methods for the calibration of measuring channels sensitivity are considered.

Since its shutdown in 1979, L-54M nuclear research reactor of Politecnico di Milano has been kept in safe storage configuration. In view of forthcoming decommissioning activities, many characterization activities have been performed on its matrices. Regarding the L-54M graphite stack, a neutron activation model was developed and validated by a graphite sampling and preliminary radiological characterization campaign focused on the most significant gamma-emitting radionuclides. Even if slightly underestimating the measured values, between − 5% and − 49% for 152Eu, between − 41% and − 67% for 154Eu, the simulated activity concentrations promisingly reproduce the experimental radial profile.

The shift from reactor to accelerator based neutron production has created a renewed interested in Boron Neutron Capture Therapy (BNCT). BNCT is reliant upon the favourable uptake of 10 B by tumour cells along with the interaction with neutrons to produce high LET fragments (He and Li nuclei) that deposit energy locally within the tumour cells. As with any radiation based treatment, Quality Assurance (QA) is crucial. In particular, Geant4 was used to model and optimise the geometry and packaging of Silicon on Insulator (SOI) microdosimeters for BNCT Quality Assurance purposes in view of experimental measurements at the KUR research reactor, in Japan. In this context, design optimisation pertains to the sensitive volume size and probability of neutron activation. This study has shown conclusively that whilst the materials currently used in the fabrication of silicon based microdosimeters are appropriate, there are changes with respect to the sensitive volume thickness that should be addressed to reduce the number of ‘stoppers’ in the microdosimeter.

The paper presents uncommon approach to the processing of neutron measurement results. The purpose of the proposed algorithm is not to restore the shape of neutron energy spectrum but to quantify the phenomena occurring in investigated neutron spectrum. The method based on minimization of functional proportional to the variance of the investigated reaction rate. The variance results from the discrepancy between the spectrum approximation and the actual spectrum. Appropriate Euler–Lagrange equation of the calculus of variations is then solved to find function minimizing the functional. The proposed algorithm is applied and tested with three different measured neutron spectra and compared with commonly used deconvolution code. The results are presented and discussed.

A rapid method for the separation and qualitative analysis of several neutron activation products (198Au, 192Ir, 72Ga, 51Cr, 191/195m/197Pt, 54Mn, 57Co, and 59Fe) from fission products and soil matrixes has been developed. Analytes were isolated within 20 h using ion exchange chromatography. After separation, the activation products were characterized by γ-spectroscopy and inductively coupled plasma-optical emission spectroscopy. Validation experiments demonstrated versatility of the method, showing that the activation products could be isolated from fresh fission products and other contaminants associated with complex soil matrixes.

In Brazil, concrete and cement are highly used in construction, therefore mixers are widely used in this industry. During the fabrication process of concrete/cement, the equipment may fail and compromise the appropriate mixing procedure. Besides that, it is also important to determine the right point of homogeneity of the mixture. It is important to have a methodology to monitor the mixing process to ensure the quality of the product. This study presents a methodology based on the principles of the radioactive particle tracking technique to predict the instantaneous positions occupied by the radioactive particle inside an industrial mixer by means of a mathematical location algorithm. The detection geometry modeled by means of MCNPX code employs an array of eight NaI(Tl) scintillator detectors, a 198Au spherical gamma-rays source with isotropic emission and a test section filled with concrete that represents an industrial mixer. The choice of the radionuclide is due its well-characterized peak of 411 keV, its half-life of 2.7 days and the possibility to obtain 198Au by neutron activation in reactors. The purpose of this study is to use an artificial neural network as a location algorithm of the 198Au radioactive particle inside an industrial mixer. Results showed that over 56% of the cases were below 5% of relative error for all coordinates of the radioactive particle, which indicates that it is possible to track the radioactive particle trajectory inside the industrial mixer using the artificial neural network algorithm.

Sulfur concentrations in soil samples containing 4.4–13.47 wt% sulphur were measured in neutron inelastic scattering using dc beams of 2.5 MeV neutrons from a DD neutron generator. The measurements were carried out using 2230 keV prompt gamma rays from sulfur using a CeBr3 detector. The minimum detectable concentration (MDC) of sulfur was 0.68 ± 0.21 wt%. The present MDC value is about one-fifth of an earlier reported value of 3.50 wt% for 3.2 MeV neutron beams. This study has shown an improvement in the sulfur MDC achieved using 2.5 MeV neutron beam.

The D-T neutron induced fission fragments from 238U, i.e., 134Te, 134 mI and 134 gI, were measured by low-background HPGe-γ detector. The fission cross-sections of the 238U(n, f) 134Te, 238U(n, f))134 mI and 238U(n, f)134 gI reactions induced by D-T neutrons, i.e., 14.7 ± 0.3, 14.5 ± 0.3 and 14.1 ± 0.3 MeV, were measured precisely with the neutron activation technique, where the neutron flux was monitored by accompanying α-particle and the neutron energy was determined by the cross-section ratio of 90Zr(n, 2 n)89Zr to 93Nb(n, 2 n)92 mNb reaction. The fission cross-sections of the 238U(n, f)134Te reaction are 54.3 ± 3.5 mb, 52.6 ± 3.4 mb and 50.0 ± 3.2 mb, corresponding to the above energy points. However, the cross-sections of 238U(n, f)134 gI reactions were estimated to be below a value about 8.2 × 10−3 mb, since almost all the 134 gI nuclide, of which characteristic γ-rays were detected in the spectrum, originated from the 134Te decay. Meanwhile, the characteristic γ-rays of 134 mI were not found on any spectrum, and beyond the lower limit of measurement.

La caracterización geoquímica de materiales arqueológicos de obsidiana o vidrio volcánico utilizado en el pasado para la confección de herramientas líticas permite trazar la procedencia geológica del mismo, y a partir de esta inferir formas de interacción y rangos de movilidad de antiguas poblaciones humanas. La caracterización de la composición geoquímica de la obsidiana por Análisis de Activación Neutrónica permite determinar con un buen grado de certidumbre su procedencia, sin embargo, es una técnica costosa y destructiva, por lo que no es recomendable para piezas arqueológicas. Otras técnicas más accesibles como la Fluorescencia de Rayos X o la Espectrometría de Masas con Plasma Acoplado Inductivamente han sido utilizadas con éxito para esta aplicación pero con un mayor nivel de incertidumbre. En este trabajo se busca asistir al problema por medio del uso de otras técnicas no destructivas que pudieran ser aplicables al patrimonio arqueológico.Método: En particular se exploraron las propiedades mecánicas de una muestra de obsidiana a través de métodos de ultrasonido. Se evaluaron la velocidad de fase y atenuación de ondas longitudinales y se cuantificó la densidad de las muestras por Picnometría de Gases, que posibilita determinar los módulos elásticos dinámicos. A su vez se repitieron las mismas mediciones en probetas de vidrio de manufactura industrial.Resultados: Se observó que la velocidad, la densidad, y el módulo longitudinal son mayores para la muestra de vidrio industrial que para la muestra de obsidiana. Por otro lado la atenuación resultó ser mayor para el caso de la obsidiana y además se observó una mayor dependencia con la frecuencia.Conclusiones: Los resultados obtenidos demuestran la capacidad del método para distinguir entre materiales con composición química diferente por lo que sería viable para ser aplicado a estudios arqueológicos. Además, las diferencias observadas en la atenuación muestran que el método también es sensible a la microestructura del material, por lo que aportaría información no accesible mediante técnicas convencionales de arqueometría. The geochemical characterization of archeological materials of obsidian or volcanic glass (used in the past for the preparation of lithic tools) allows to trace origin and infer forms of interaction and ranges of mobility of ancient human populations. The characterization of the geochemical composition of obsidian by analysis of Neutron Activation allows to determine with a good degree of certainty its origin; however, it is a costly and destructive technique, not recommended for archeological pieces. Other more accessible techniques, such as X Ray Fluorescence or Mass Spectrometry with inductively coupled plasma have been used successfully for this application but with a higher level of uncertainty. This work seeks to provide solutions through the use other non-destructive techniques that may be applicable to archeological heritage. Fil: Cowes, Diego Alejandro. Universidad Nacional de Tres de Febrero; Argentina. Comisión Nacional de Energía Atómica; Argentina Fil: Salgán, María Laura. Consejo Nacional de Investigaciones Científicas y Técnicas. Centro Científico Tecnológico Conicet - Mendoza. Instituto Argentino de Nivología, Glaciología y Ciencias Ambientales. Museo de Historia Natural de San Rafael - Ianigla | Provincia de Mendoza. Instituto Argentino de Nivología, Glaciología y Ciencias Ambientales. Museo de Historia Natural de San Rafael - Ianigla | Universidad Nacional de Cuyo. Instituto Argentino de Nivología, Glaciología y Ciencias Ambientales. Museo de Historia Natural de San Rafael - Ianigla; Argentina. Universidad Tecnológica Nacional. Facultad Regional de San Rafael; Argentina Fil: Mieza, Juan Ignacio. Universidad Nacional de Tres de Febrero; Argentina. Universidad Nacional de San Martín; Argentina. Comisión Nacional de Energía Atómica; Argentina. Consejo Nacional de Investigaciones Científicas y Técnicas; Argentina Fil: Gómez, Martín Pedro. Universidad Nacional de San Martín; Argentina. Universidad Tecnológica Nacional. Facultad Regional Delta; Argentina. Comisión Nacional de Energía Atómica; Argentina

Recently the incidence rate of renal stone disease is increased in Egypt. In addition to industrial and environmental effects in this increase, trace elements may also have a role in the formation of such stones, inspit of their significance in biological lithogenesis. In this study instrumental neutron activation analysis (INAA) technique is applied for elemental analysis of some renal stones, collected from patient living in different provinces in Egypt. The samples are collected from male, female and children. The patients are of varied ages. The analyses are carried out under various conditions of irradiation and cooling time. Major, minor and trace elements were determine. Standard material is used for accuracy and quality control measurements. The analyzed elements were: As, Br, Ca, Co, Cr, Cu, Fe, Hf, In, Mg, Mn, Na, Se, Sr, Tb, V and Zn under the experimental conditions Some of these elements are biologicaly essential and others are dangerous or toxic.

To ensure that fuel rods operate in nuclear reactors safely and reliably, UO2 pellets with different enrichment levels of 235U in the same production line are manufactured in batches and divisionally managed to avoid confusion or the potential misloading of UO2 pellets with different enrichment levels. At the same time, nondestructive tests for their enrichment levels and loading uniformity and all UO2 pellets must be nondestructively tested during production. By studying the enrichment detection mechanism of the UO2 pellets of 235U, the design of an integral standard rod was carried out, and a single integral standard rod was used to achieve the calibration of the enrichment measurement curve, as well as the detection and calibration of abnormal pellets. This study undertook a comparison test of 235U enrichment between the neutron activation method and the array multi-probe passive method. The test results showed that the array multi-probe passive method had higher detection efficiency and equal accuracy.

The neutron activation cross section for $$^{127}$$ I(n, $$\gamma $$ ) $$^{128}$$ I reaction has been experimentally measured with respect to the $$^{115}$$ In(n,inl) $$^{115}$$ In $$^{m}$$ reference monitor reaction cross section in the neutron spectrum average energy range 0.60 $$--$$ 2.51 MeV. The neutrons were produced through $$^{7}$$ Li(p,n) $$^{7}$$ Be reaction, and $$\gamma $$ -ray spectra of the residue product were measured offline with the precalibrated lead-shielded HPGe detector. The very first time the covariance analysis was done to quantify the measured cross section uncertainties and the correlation coefficients between the different neutron energy cross sections for iodine nucleus. The needful corrections related to the $$\gamma $$ -ray self-attenuation process, $$\gamma $$ -ray true coincidence summing effect and the low background neutron energy contributions were considered in the present measurement. Theoretical calculations were done using the standard nuclear reaction model codes TALYS-1.9 and EMPIRE-3.2 to obtain the $$^{127}$$ I(n, $$\gamma $$ ) $$^{128}$$ I reaction cross section. The measured cross sections were compared with the experimental data available in the EXFOR database, theoretical predicted results, and ENDF/B-VIII.0, JEFF-3.1/A, TENDL-2019, and JENDL-4.0 evaluated nuclear data.

The cross-section function for the $$^{115}$$ In( $$\gamma $$ , $$\gamma ^{\prime }$$ ) $$^{115m}$$ In reaction was determined in the energy range up to E $$_\gamma $$ = 9.6 MeV using the bremsstrahlung facility at the MT25 Microtron, JINR, Dubna. Natural indium disks were irradiated with bremsstrahlung, each disk with a beam of different endpoint energy. To the measured saturation activity built up in the wide-energy photon beam we applied the unfolding technique, developed recently at the JRC-Geel with the primary purpose of being used in neutron activation. The results were compared with TALYS 1.9 calculations and existing experimental data. Our results suggest that the application of existing unfolding technique allow determining unknown excitation functions of photon-induced reactions.

Neutron Activation Analysis (NAA) is a well-established nondestructive analytic technique where the gamma radiation emitted by an irradiated sample is analyzed using an HPGe detector. The Neutron Activation Laboratory (LAN) of IPEN-CNEN/SP has been performing NAA analyses for over 30 years, and has plans of implementing quality control protocols to their analyses. In this sense, the environmental monitoring of the laboratories where the detectors are used has been performed for many years, in a manual way with no more than 2 measurements per day. In this work, an automated monitoring station based on a microcontroller Arduino UNO board has been developed which comprises four thermo hygrometer sensors for monitoring different parts of the environment, plus a thermocouple for monitoring the inside of the liquid nitrogen dewar. The results obtained allow for a discussion on the performance and adequacy of the sensors.

The cross sections of the selenium isotopes 76Se, 77Se, 78Se and 80Se within 10.5–19.81 MeV neutron energy range have been measured through neutron activation method along with off-line γ-ray spectrometry. The quasi-monoenergetic neutrons were produced from the 7Li $$\left( {p, n} \right)$$ reaction at 14UD BARC-TIFR Pelletron Accelerator Facility, Mumbai, India. The statistical codes TALYS-1.9 and EMPIRE-3.2.2 were applied for the theoretical calculation of reaction cross sections with different level density models from 2 to 22 MeV neutron energies. Besides this, the Se $$\left( {n, p} \right)$$ As reaction cross sections were also calculated from different systematic formulae within 14–15 MeV neutron energies. The measured data were compared with existing literature data available in the EXFOR database, evaluated data of ENDF/B-VIII.0, JENDL-4.0 and TENDL-2019 libraries and with theoretical outcomes through TALYS-1.9 and EMPIRE-3.2.2 codes. The uncertainties in existing cross sections were calculated through the method of covariance analysis by including partial uncertainties and correlation among the different attributes. The $$\left( {n, p} \right)$$ reaction cross sections of selenium isotopes at higher neutron energies first time measured in the present work can be added as new data in the nuclear data library.

Fil: Ratto, Norma Rosa. Consejo Nacional de Investigaciones Científicas y Técnicas. Oficina de Coordinación Administrativa Saavedra 15. Instituto de las Culturas. Universidad de Buenos Aires. Instituto de las Culturas; Argentina. Fil: De la Fuente, Guillermo Adrián. Consejo Nacional de Investigaciones Científicas y Técnicas. Centro de Investigaciones y Transferencia de Catamarca. Universidad Nacional de Catamarca. Centro de Investigaciones y Transferencia de Catamarca; Argentina. Universidad Nacional de Catamarca. Escuela de Arqueología. Laboratorio de Petrología y Conservación Cerámica; Argentina. Fil: Vera, Sergio David. Consejo Nacional de Investigaciones Científicas y Técnicas. Centro de Investigaciones y Transferencia de Catamarca. Universidad Nacional de Catamarca. Centro de Investigaciones y Transferencia de Catamarca; Argentina. Universidad Nacional de Catamarca. Escuela de Arqueología. Laboratorio de Petrología y Conservación Cerámica; Argentina. Fil: Feely, Anabel. Consejo Nacional de Investigaciones Científicas y Técnicas. Oficina de Coordinación Administrativa Saavedra 15. Instituto de las Culturas. Universidad de Buenos Aires. Instituto de las Culturas; Argentina. Fil: Callegari, Adriana Beatriz. Universidad de Buenos Aires. Facultad de Filosofía y Letras. Instituto de Arqueología; Argentina. Fil: Plá, Rita Rosa. Comisión Nacional de Energía Atómica. Gerencia del Área de Energía Nuclear. Departamento de Química Nuclear; Argentina. Pottery production during the Early Period at Northwestern Argentine region has been characterized by the existence of a Grey-Horizon involving vessels fired under extreme reducing conditions, probably using some kiln-based firing structures. Neutron activation (NAA) and technological analysis were carried out on the Early Period pottery from 12 archaeological sites geographically located mostly at southwestern Catamarca and northern La Rioja provinces. Ceramic fabrics analysis suggests that several technological choices were implemented by potters as different regional potteŕs recipes to produce the vessels. Geochemical groups as defined by NAA indicate that at least three local production areas were working during the Early Period in these two geographical areas. Potters used local clays to produce most types of vessel forms.

The fast and thermal neutron attenuation properties through polymer composites based on high density polyethylene (HDPE) reinforced by micro-sized and nano-sized Cadmium Oxide (CdO) particles with weight fractions of 10% and 40% have been investigated. This study was carried out to present a new composite material based on HDPE filled with CdO in the form of micro and nano particles to be used as a promising neutron radiation shielding material. The composites were fabricated by compression molding technique and characterized by a scanning electron microscope (SEM). The composites were subjected to fast neutrons generated from 238Pu–Be neutron source with activity 12 × 109 Bq and detected by a Stilbene scintillator. The fast neutron shielding properties were determined in terms of the fast neutron transmission fractions of the composites. The neutron activation method was applied by using Indium foils and Indium foils covered with Cadmium as a detector to study the properties of the thermal and epithermal neutron fluxes of the composites upon exposure to neutron flux generated from 241Am–Be neutron source. It is clear from this study that the transmitted fractions for fast neutron in case of nano CdO/HDPE composites are less than the case of micro CdO/HDPE composites, however, the capture of thermal neutrons inside the micro CdO/HDPE composites is more probable than that in nano CdO/HDPE composites at the same weight fraction. Tensile mechanical testing has been conducted and showed that particle addition percentage and size have significant effect on the mechanical properties of composites. Nano CdO/HDPE composites showed more superior mechanical properties compared to micro CdO/HPDE composites at the same particle addition level.