Neutron Activation Research Articles

Measurement of tritium production in the helium cooled pebble bed test blanket module mock-up at JET during DTE2

Quite often, detectors for measuring nuclear performance and radiation quantities of relevance in fusion experiments are requested to withstand harsh working conditions due to intense neutron and gamma radiation fields. High temperature constitutes a further harsh element in some locations of the machine, where it is necessary to perform some on-line measurements, as expected in the breeding blanket. This is an essential component in future fusion power plants to provide tritium self-sufficiency and its performance must be continuously monitored. Some Test Blanket Modules (TBMs) will be installed in ITER to provide the first experimental data to validate the predictions on tritium production and recovery. In the meantime, within EUROfusion program, the mock-up of the Helium Cooled Pebble Bed Test Blanket Module (HCPB TBM), previously used for the TBM experiment at the Frascati Neutron Generator (FNG), had been installed at JET to test some detectors and for benchmarking numerical codes used for breeding blanket assessment during DTE2 campaign. A diamond detector, calibrated to measure the tritium production through neutron detection inside the HCPB TBM mock-up, was tested during some plasma pulses of the DTE2 campaign at JET. The main outcome is that, as far as neutron emission rate is below 1015 s−1, neutrons are properly detected along the plasma discharge evolution by TBM diamond detector, consistently with the JET neutron monitor KN1. Moreover, the amount of tritium measured (E) is 1.40 × 10–12 tritons per source neutron and the comparison with MCNP radiation transport simulation (C) gives a ratio C/E = 0.77. Such measurements, considered promising, and their comparison with calculations are discussed in the present work. Criticalities emerged are analyzed and some improvements proposed with the main purpose of speeding up signal processing to make the system capable of working at higher plasma neutron emission rates.

Impact of neutron radiation induced defects on the surge current robustness of silicon carbide P‐intrinsic‐N diodes

AbstractSurge current (SC) capability is one of the main aspects of reliability for silicon carbide (SiC) power devices. In this work, the influences of neutron radiation‐induced defects on the SC capability and reliability of SiC P‐intrinsic‐N (PiN) diodes were comprehensively investigated. It was found that the surge capability of the diodes can be deteriorated even under the slightly enhanced formation of carbon‐vacancy‐related Z1/2 and EH6/7 defects introduced by neutron irradiation. Surprisingly, it was found that the forward voltage (VF) decreases with the increased SC and the stress cycles in the irradiated diodes, which is usually found to increase under the SC tests and attributed to the bipolar degradation (BPD). By using technology computer‐aided design simulation and deep‐level transient spectroscopy characterization, it was found that the significant self‐heating during surge stress leads to the annealing effect on the Z1/2 defects through the promoted recombination with the nearest and second neighbor carbon interstitials injected by irradiation, which thus plays a dominant role in the decrease of VF over the BPD.

Application of instrumental neutron activation analysis (INAA) to ceramics and clay sources for the late period (c. 950 years BP) in the Southern Argentine Puna: Interweaving sources, objects and practices

Application of instrumental neutron activation analysis (INAA) to ceramics and clay sources for the late period (<i>c</i>. 950 years BP) in the Southern Argentine Puna: Interweaving sources, objects and practices

Bout the composition and neutron activation radiography of the topaz crystals from the author’s collection

Bout the composition and neutron activation radiography of the topaz crystals from the author’s collection

A comparative investigation of neutron and gamma radiation interaction properties of zircaloy-2 and zircaloy-4 with consideration of mechanical properties

Abstract This study has established the radiation shielding efficacy of zircaloy-2 and zircaloy-4 over a wide spectrum of energy levels. Using the Monte Carlo N-Particle (MCNP) method, the gamma and neutron transmission factors (TF and nTF) were calculated for various energy levels. Zircaloy-2 demonstrated the highest gamma-ray absorption capacity and the lowest neutron absorption capacity among the investigated alloys. The results indicate that zircaloy-2 and zircaloy-4 have nearly the same neutron transmission characteristics. Although many studies have examined the structure and physical characteristics of these materials, there has been a lack of Monte Carlo simulations to comprehensively investigate the correlation between gamma absorption, neutron absorption parameters, and mechanical qualities. This research aims to examine the ability of zirconium and its zircaloy-2 and zircaloy-4 alloys, which are critical materials used in the nuclear industry, to absorb gamma and neutron radiation over a broad spectrum of frequencies. According to the results, zircaloy-2 has the best ability to absorb secondary gamma rays and the highest level of resistance to them. Despite the minimal disparity in the nTF between the two alloys, simulation results have shown that zircaloy-2 has a higher level of neutron transmittance. These results have the potential to expedite the development of novel materials with enhanced attributes for various applications.

Radiation shielding properties of sustainable concrete with novel plastering techniques

In concrete applications. Major/critical applications of such concrete are radiation-shielding facilities. Both steel slag and silica fume are examples of common by-product materials that can be used as a replacer of aggregates and cement. Thus, in this research work, steel slag was utilized as heavy aggregate in concrete production besides silica fume to present sustainable concrete mixtures probably with better radiation-shielding properties. Different cementitious plasters were applied on the conducted sustainable concrete mixture using different powdery materials; hematite, magnetite, barite, bentonite, and steel slag powders in addition to nano-titanium dioxide as full replacers for sand. The proposed plasters were presented to determine the optimum plaster technique in terms of static performance and attenuation capability against gamma and neutron radiations. The results exhibited that utilizing steel slag and silica fume in concrete mixtures enhanced compressive strength by up to 9.09 % compared to conventional concrete, while the addition of nano-titanium to conventional plaster led to superior enhancement in the compressive strength by up to 38.65 % relative to traditional plaster. Conversely, fully replacing conventional silica sand with the abovementioned powdery materials generally reduced the compressive strength of cementitious plasters by up to 30.83 %. However, the radiation shielding properties against Cs-137, and Co-60 energies have been enhanced by up to 20 % and 26 %, respectively.

Development of a thermal diagnostic system for the SPARC tokamak.

The SPARC tokamak will have scrape-off layer parallel heat fluxes on the order of GW/m2. Managing power exhaust of this magnitude will be mandatory for a reactor-scale device. To enable this mission, a thermal diagnostic suite will be deployed to measure the in-vessel structural temperatures to ensure they do not exceed their design limits and to determine the spatial distribution and magnitude of energy deposited onto the first wall. Thermocouples and fiber Bragg gratings have been selected for their environmental compatibility and proven useful on other fusion devices. High-density thermocouple arrays in the divertor will have two spring-loaded thermocouples per divertor target tile, which are being used as calorimeters, and will look to resolve the temperature distribution within the tile due to a swept or static strike point. All systems will need to survive the vacuum vessel bake, set at a minimum plasma facing surface temperature of 350 °C, which presents a particularly challenging environment for the fiber-based subsystem. Along with this temperature design requirement, all the materials in the primary vacuum need to be ultra-high vacuum compatible, able to handle the expected neutron and gamma radiation, as well as tritium exposure, all of which restrict material options. Finally, due to the expected activated environment in SPARC, there will be little chance to replace defective sensors, so system resilience is ensured through toroidal redundancy, probe material selection, and mitigating the impact of common-mode failures. Initial testing of sensors show that intershot structural measurements are sufficiently captured with the raw output, but intrashot measurements of the plasma facing material requires model-based interpretive tools.

Commissioning of the MIGDAL detector with fast neutrons at NILE/ISIS

Many dark matter experiments are exploiting the Migdal effect, a rare atomic process, to improve sensitivity to low-mass (sub-GeV) WIMP-like dark matter candidates. However, this process is yet to be directly observed in nuclear scattering. The MIGDAL experiment aims to make the first unambiguous measurement of the Migdal effect in nuclear scattering. A low-pressure optical Time Projection Chamber is used to image in three-dimensions the characteristic signature of a Migdal event: an electron and a nuclear recoil track sharing a common vertex. Nuclear recoils are induced using fast neutrons from a D-D source, which scatter in the gaseous volume of the detector. The experiment is operated with 50 Torr of CF4 using two glass GEMs for charge amplification. Both light and charge are read-out, and these measurements are combined for track reconstruction. Commissioning data has been recorded with fast neutrons at the Neutron Irradiation Laboratory for Electronics (NILE) at Rutherford Appleton Laboratory in the UK. Results of the experiment’s commissioning and the performance of the detector with a high rate of highly ionising nuclear recoils are presented, along with results from low energy electrons. Initial results of light and charge read-out with low pressure noble gas mixtures are also presented.

Preliminary study of a compact epithermal neutron absolute flux intensity measurement system for real-time in-vivo dose monitoring in boron neutron capture therapy

Boron neutron capture therapy (BNCT) is a radiotherapy technology that selectively kills tumor cells via the 10B(n, α)7Li reactions. Epithermal neutrons (0.5 eV to 10 keV) are emitted and converted into thermal neutrons, which have a larger neutron capture reaction cross-section, by slowing down in the human body before reaching the tumor. Recently, the development of an epithermal neutron absolute flux intensity measurement technique has become crucial for real-time in-vivo dose monitoring in BNCT. In this study, a concept for a measurement system consisting of multiple compact scintillator with optical fibers detectors covered with neutron absorbers of various thicknesses is proposed. The designed system achieves a consistent response to epithermal neutrons with a theoretical coefficient of variation not higher than 5% for both LiCAF and EJ-254 scintillators. The theoretical feasibility of the proposed measurement system was investigated by an irradiation experiment carried out at the heavy water neutron irradiation facility at the Kyoto University Reactor. The experimental results indicated that further improvement and refinement are necessary to meet the high accuracy and precision required for real-time dose monitoring in clinical applications.

Experimental and simulation studies to optimize the mechanical alloying process of ODS-FeCrAl alloy powder

Oxide dispersion strengthened (ODS) FeCrAl alloy exhibits superior mechanical qualities at high temperatures and resistance to neutron irradiation, making it a highly promising material for accident-tolerant fuel cladding. The preparation of ODS-FeCrAl alloy powder by mechanical alloying can mix Y2O3 and FeCrAl alloy powder uniformly and promote the uniform dispersion of nanoscale oxide particles in the alloy matrix. The preparation of ODS-FeCrAl alloy powder by the mechanical alloying method is greatly influenced by the rotational speed. Therefore, this study uses a combination of experiments and numerical simulation to investigate the effects of different rotational speeds on the microscopic morphology, particle size and force of alloy powder during the ball milling process. The mechanically alloyed powder was subjected to XRD inspection to determine the optimum ball milling time at each rotational speed by means of the maximum lattice distortion, and then the powder micromorphology and particle size were analyzed. Then the forces on the powder and grinding ball during the ball milling process were analyzed by numerical simulation, and it was found that the collision force and normal force played a dominant role in the ball milling process. The combined experimental and simulation results determined that the ODS-FeCrAl alloy powder obtained by ball milling at 300 rpm for 35 h had the largest lattice distortion, fine grain size and uniform particle size, which achieved the best mechanical alloying effect.

Predicting scale thickness in three-phase flow using neutron activation analysis and deep learning

Calcium carbonate scales occur on the inner walls of the pipes used in oil and gas production, reducing the internal diameter and obstructing fluid flow. This study proposes a method for predicting of scale thickness using a neutron source and gamma radiation analysis with a deep neural network (DNN). The detection system includes a 241Am-Be neutron source and 72 detectors that count particles and record photons. A measurement setup was designed using the MCNP6 code to optimize the detection of scale thickness, utilizing a flat neutron source and strategically positioned detectors. The gamma-ray spectra generated from this setup were used as input data for the DNN to predict scale thickness within a three-phase water–gas-oil annular system. The results indicated that the DNN accurately predicted scale thickness with a mean relative error of 3.01 %, achieving a relative error below 5 % for 92.68 % of the dataset, demonstrating the potential for reliable real-time monitoring in industrial settings.

Boron neutron capture therapy and environmental health

Boron Neutron Capture Therapy (BNCT) is an innovative cancer treatment modality that exploits neutron beams to specifically destroy malignant cells. Despite its assuring clinical benefits, BNCT familiarizes itself with unique environmental health considerations, including radiation safety, waste supervision, and potential ecological ramifications. This review investigates these aspects by appraising radiation safety protocols, waste disposal procedures, and environmental monitoring implementations associated with BNCT facilities. Effective radiation shielding and demanding safety protocols are necessary to protect occupational and public health from potential neutron radiation exposure. Waste management implementations must inscribe the treatment, handling, and disposal of radioactive and biological waste to anticipate environmental contamination. Long-term ecological monitoring is decisive for detecting and mitigating any adverse influences on surrounding ecosystems. The review also emphasizes the significance of public awareness and education in addressing environmental health concerns. By integrating case studies from various international perspectives and recognizing research gaps, this review aims to contribute comprehensive insights into the ecological health suggestions of BNCT and propose strategies for enhancing its sustainability. The findings accentuate the need for continuous improvement in safety measures, waste management, and public engagement to corroborate that BNCT can be safely and effectively implemented in a technique that protects both human health and the environment.

Measurement of fast neutron induced (n,γ) reaction cross-section of 68Zn, 96Zr, 121Sb and 123Sb in the energy range of 1 to 2 MeV

The (n,γ) reaction cross-section for the elements 68Zn, 96Zr, 121Sb and 123Sb, present in the reactor structural/shielding materials, was measured by neutron activation technique in the neutron energy region of 1–2 MeV as very limited data is available in this energy range. Further, the neutron spectrum peaks in this energy region for the fast breeder reactors and proposed accelerator driven sub-critical systems. The natural strontium (natSr) element was used as a neutron flux monitor by considering effective combined reaction cross-section for 86Sr(n,γ)87Srm and 87Sr(n,n′)87Srm reactions. The pellets of mixture of sample and monitor were irradiated by a quasi-mono energetic fast neutron beam, generated by 7Li(p,n)7Be reaction at FOTIA, Bhabha Atomic Research Centre, Mumbai, India. The activity of activation products was measured by off-line gamma-ray spectrometry using High Purity Germanium Detector (HPGe). The present data with improved uncertainty and covariance analysis enhance the cross-section data base for better constraining the evaluated data and theoretical models. The theoretical (n,γ) reaction cross-sections were calculated using TALYS 1.96, which could reasonably explain the present data with the Fermi gas level density prescription.

Characterization of neutron-irradiated SiPMs down to liquid nitrogen temperature

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The SiPMs were also annealed at high temperatures and re-characterized after the annealing. For the SiPM characterization in all the cases, current–voltage (I-V curve) measurements, dark count rate measurements, and waveform analysis, including single photon time resolution, were carried out at different controlled temperature steps from room temperature down to the liquid nitrogen temperature. 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With sufficient cooling, the measured single photon time resolution was around 90 ps FWHM for all irradiation levels.

On neutron generator applications in Brazil: current panorama and perspectives

Techniques based on neutron beams are used in research, industry and medicine being especially suitable for the characterization of a wide variety of materials. Neutron radiation can be produced using nuclear reactors, isotopic sources, or particle accelerators. Since the number of reactors is in decline and isotopic sources are expensive and difficult to handle, neutron generator (NG) technology has experienced significant development. Neutron generators are safer than nuclear reactors and isotopic sources, and the use of NGs is increasing worldwide, including in Brazil. This article reviews the main applications of neutron radiation, neutron production techniques, and specifically the technologies used in neutron generators. An overview of the utilization of these techniques in Brazil is presented, along with studies on acquisition and start-up costs of neutron-generating equipment and perspectives for future utilization in various areas.

Degradation of electrical properties of subcells in multi-junction solar cells under neutron irradiation

This paper presents an analysis of the photovoltaic characteristics and parameters of individual subcells of space multi-junction solar cells after irradiation by high-energy particles. Dark currents, charge carrier lifetimes, and damage coefficients for wide-bandgap subcells were determined both theoretically and experimentally.

Polymer Boron-Containing Composite for Protecting Astronauts of Manned Orbital Stations from Secondary Neutron Radiation

This article considers the prospects of using heat-resistant polyimide boron-containing composites to protect astronauts of manned orbital stations from secondary neutron radiation. Variant calculations are performed regarding neutron and gamma-quanta flux distributions in a polyimide composite material with different boron content used to reduce capture radiation. The dependences of spatial distributions of thermal neutron flux density and the gamma-quanta dose rate in a polyimide composite layer with a boron content of 0 to 5% are obtained. An experimental assessment of the energy distribution of neutron and gamma radiation behind the protective polyimide composite is carried out. The introduction of boron atoms in an amount of 3.0 wt.% shows the absence of bursts of secondary gamma radiation energy in the composite, which is due to the high cross-section of thermal neutron absorption by boron atoms. As a result, with a material layer thickness of 3–10 cm, the gamma-quanta dose rate decreases by 2–3 times. The differential thermal analysis method showed that the upper limit of the working temperature of the polyimide composite is 500 °C. The polyimide matrix filled with boron atoms can find effective application in the development of new radiation-protective polymer materials used in manned orbital stations.

Determination of the silicon content of geological materials by reactor fast neutron activation analysis

Determination of the silicon content of geological materials by reactor fast neutron activation analysis

Effect of grain boundaries on the helium degradation mechanisms of alloy 800H: A molecular dynamics study

Alloy 800H is currently used as structural material in light-water cooled nuclear reactors, and it is also considered as candidate materials for various advanced reactor designs. Under operation, the exposure of alloy 800H to neutron irradiation results in the formation of helium (He), mainly from Ni by the transmutation reaction (n, α). In this work, we present a molecular dynamics (MD) simulation study of the behavior and effects of He on the microstructure and mechanical properties of alloy 800H. Our results show that the population of clusters made of 5 to 9 He atoms is nearly constant throughout the simulation time (10 ns), while the population of larger clusters increases as the simulation time increases. The growth of clusters is controlled by either the dissociation and diffusion of smaller clusters towards nearby larger clusters or the merging of larger clusters initially located nearby to each other. A significant accumulation of He is observed at the grain boundaries (GB), while a depletion zone is found at the neighboring regions. As a result, the density of He cluster is significantly higher at the GBs as compared to the intra-granular regions. The nucleation and growth of He clusters also results in the formation of Frenkel pairs (FP), whose associated self-interstitial atoms (SIA) agglomerate into interstitial clusters in the alloy 800H matrix. As a consequence, dislocation segments, mostly of the Shockley type, are generated in the microstructure, and often located next to He clusters. The combination of the aforementioned defect structures and the high density of He clusters at the GBs results in a substantial degradation of the mechanical properties of alloy 800H single crystal and bicrystals.

Radiological protection of Young Scientist and Scientific Researcher trainees in controlled areas of the Argonaut reactor installation: analytical approach

Since the increasing use of ionizing radiations in medical and industrial proceedings, recommendations and actions have been necessaries to reduce the individual radiological risks. In the Nuclear Engineering Institute (IEN), there is the Argonauta research nuclear reactor (RARG) used for producing radionuclides, neutrongraphy studies and training young scientists (YST) and scientific researchers (SRT). During their trainings, these trainees are exposed to ionizing radiations meanly during the neutrongraphyc experiments when they need to adjust neutrongraphy plates in front the J9 channel with the RARG in critical condition. At this moment, they are exposed to neutron and gamma radiations in the RARG hall. This paper evaluated the gamma and neutron expositions that YST and SRT are submitted during the neutrongraphy experiments, by using area monitors to measure the ambient dose equivalent rates in two points near to J9 channel. Considering only this experiment, there was found that the annual individual effective dose could reach up to 2.5 mSv, which is more than 150% of the annual dose to member of the public (MP), but below the annual effective dose to occupationally exposed individual (OEI) as a worker. The question is if YST and SRT should be considered an OEI or a MP. This question is answered through an analytical approach of the regulations and recommendations. The conclusion is that YST and SRT shall be controlled as OEI, but must be submitted to MP’ restriction doses.