High Neutron Yield Research Articles

Preliminary evaluation of alloying significance of neutron target for high brilliance neutron source

Accelerator-based neutron sources are critical infrastructures in today's scientific research and industrial fields, with the neutron source target being the key component determining the performance of the neutron source. Traditional accelerator neutron source targets typically use homogeneous elemental materials with high neutron yield as the target material. These materials face challenges in terms of operational lifespan under high-temperature and high-radiation environments. The design of next-generation high brilliance neutron sources demands target materials with better thermal conductivity, longer irradiation lifespan, and superior corrosion resistance. Tungsten has its high neutron yield, and tungsten alloys exhibit these advantageous properties. This paper employs the Monte Carlo method to calculate neutron yield, radioactivity and radiation damage of two common tungsten alloys, W-TiC and W-ZrC. The results indicate that, with minor doping to improve mechanical properties and manufacturability, the performance of tungsten alloys can also meet the requirements for high-brilliance neutron targets, provide reference for the development of the next generation neutron target.

Neutron-producing gas puff Z-pinch experiments on a fast, low-impedance, 0.5 MA linear transformer driver

A study on the neutron production from single and double gas puff Z-pinches on the CESZAR linear transformer driver with ∼0.45 MA current and 170 ns rise time is presented. Total neutron yield measurements made with a LaBr activation detector are compared for three configurations, using a double nozzle setup. When a single, hollow, deuterium gas shell was used, reliable implosions could only be attained at higher load mass than the optimal value to match implosion time with the driver rise time, with neutron yields of ∼106 per pulse. The use of a double gas puff configuration with a deuterium center jet allowed a reduction in the shell density and operation closer to machine-matched conditions, recording up to (4.1 ± 0.3) × 107 neutrons/pulse when either Kr or D2 was used in the shell. For a comparable mass and implosion time, using a higher atomic-number gas in the outer shell results in more unstable plasma surface and smaller plasma radius at the location of instability bubbles, which, however, do not seem to consistently correlate with a higher neutron yield. Comparing implosion dynamics with models and neutron yields with literature scaling suggests that the machine current is not well coupled to the plasma during the final stages of compression. Optimizing current and energy coupling to the pinched plasma is critical to improving performance, particularly in low-impedance drivers.

Variable-sagittal-radii elliptical x-ray crystal spectrometers for high-neutron-yield plasma diagnostics.

Sagittally focusing x-ray crystal spectrometers with elliptical profiles in the meridional (x-ray dispersion) plane are proposed for plasma diagnostics in experiments accompanied by high neutron yields. The spectrometers feature a variable sagittal radius of curvature to ensure the sagittal focusing of rays for each photon energy in a chosen detection plane. The detector is placed after the ray crossing point at the second ellipse focus, and the source-to-detector distance is maximized to reduce the neutron-induced background. The elliptical shape imposes a limitation on the spectrometer geometry such that the influence of the source size on the spectral resolution can be avoided only for a demagnifying spectrometer (the source-to-crystal distance is larger than that of crystal-to-detector). Hence, two designs are proposed. The first design, featuring high magnification and limited spectral resolution can be suitable for x-ray continuum spectroscopy. The second design of high demagnification is optimized for spectral resolution, and can be used for time-resolved spectroscopy of plasma's characteristic emission lines using streak cameras. The key performance characteristics of the two designs are verified using ray tracing.

Characterization of a CMOS camera based film digitization platform for gated x-ray imaging diagnostics at the National Ignition Facility.

Hardened gated x-ray detectors use photographic film as the data recording medium due to its low sensitivity to the high-yield neutron environments at the National Ignition Facility (NIF). The photographic film is digitized with a Photometric Data Systems (PDS) microdensitometer, which measures the film's optical density. The PDS scanner is able to measure a dynamic range of 0-5 OD; however, raster scanning the film is time consuming and maintenance of the instrument is challenging due to legacy technology. Since film usage at NIF is expected to continue in the foreseeable future, a digitization platform that is faster and more maintainable would benefit the NIF's current and future operations. This work presents the characterization of the digital transitions (DT) atom, a CMOS camera-based digitization platform that records film data in a single image capture very quickly and has widely available user support. The preliminary results suggest that the DT atom is able to reconstruct exposures accurately enough to be a competitive alternative to the PDS Scanner.

Mitigation strategies for time-resolved x-ray diagnostic data degradation due to high neutron yield of ICF experiments at the NIF.

Inertial confinement fusion experiments taking place at the National Ignition Facility are generating ever increasing amounts of fusion energy, with the deuterium tritium fusion neutron yield growing a hundredfold over the past ten years. Strategies must be developed to mitigate this harsh environment's deleterious effects on the operation and the performance of the time-resolved x-ray imagers deployed in the National Ignition Facility target bay to record the dynamics of the implosions. We review the evolution of these imagers in recent years and detail some of the past and present efforts undertaken to maintain or improve the quality of the experimental data collected on high neutron yield experiments. These include the use of a dump-and-read electronic backend, the selection of photographic film with a low background sensitivity, and the optical filtering of Cherenkov radiation.

A protype of compact and high-yield DD neutron generator

Neutron generators are widely utilized in various fields, including scientific research, activation analysis, and detection of IED or drugs, due to their high neutron yield, safety, and flexibility. In this paper, we report about the development of a compact and high-yield dd neutron generator based on an Electron Cyclotron Resonance (ECR) ion source at the China Institute of Atomic Energy. The ion source to produce a high-intensity ion beam, which is used to bombard a water-cooled solid metal target. By means of the (D, D) nuclear reaction, 2.5MeV neutrons are generated. We describe the detailed design of this compact dd neutron generator and present the preliminary measurement results of its neutron yield, as measured by a calibrated 3He neutron detector. The generator has been tested for nearly 300 hours, demonstrating stable operation without any component damage. With its compact size, high yield, and long lifespan, this generator could meet the demand for a 2.5MeV neutron source in many scientific research and select industrial applications.

Neutron Activation Analysis Based on AB-BNCT Treatment Room.

Boron neutron capture therapy (BNCT) is an ideal binary targeted radiotherapy for treating refractory tumors. An accelerator-based BNCT (AB-BNCT) neutron source has attracted more and more attention due to its advantages such as higher neutron yield in the keV energy region, less gamma radiation, and higher safety. In addition to 10 B, neutrons also react with other elements in the treatment room during BNCT to produce many activation products. Due to the long half-life of some activation products, there will be residual radiation after the end of treatment and the shutdown of the accelerator, which has adverse effects on radiation workers. Therefore, the ambient dose equivalent rate in the treatment room needs to be evaluated. The AB-BNCT neutron source model proposed by Li is studied in this paper. Based on the Monte Carlo method, the Geant4 platform was used to simulate the dose induced by radionuclides near the Beam Shaping Assembly (BSA) of the source. It is concluded that the concrete wall contributed the most to the radiation dose. The dose rate of 2.45 μSv h -1 after 13 min of shutdown meets the dose rate limit of 2.5 μSv h -1 , at which point it is safe for workers to enter the treatment room area.

Study of accelerator-based epithermal neutron flux depending on Li and Be targets

The Li(p,n) and Be(p,n) reactions are widely used as accelerator-based neutron sources for boron neutron capture therapy (BNCT). Because the energy of the neutrons produced by these reactions is high for BNCT, a moderator is required to reduce the energy to the epithermal energy region. The neutron yields and energy spectra derived from the Li(p,n) and Be(p,n) reactions vary with the incident proton energy; a high proton energy results in a high neutron yield. However, a high proton energy can also increase the neutron energy, and in this case, a high neutron energy requires a thick moderator, which decreases the epithermal neutron flux. Notably, these tradeoffs are not completely clear. Furthermore, the upper limit of the epithermal neutron energy used in different BCNT facilities is not the same. Therefore, epithermal neutron flux estimation with applying an appropriate moderator is necessary for comparing epithernal neutron flux depending on the proton energy, target type, and epithermal neutron range definition.In this study, GEANT4 simulations were performed to determine effective moderator materials and thicknesses. After applying a moderator, we examined the epithermal neutron flux by changing the conditions that can influence the epithermal neutron flux, such as the proton energy, target species (Li and Be), and different definitions of the epithermal neutron region. The results showed that the epithermal neutron flux increases with the increasing proton energy until a certain level for both Li and Be targets. For proton energies above 3 MeV for the Li target and above 7 MeV for the Be target, the epithermal neutron flux decreases or its increase rate gradually decreases. And, when the upper energy limit of the epithermal neutron range is changed from 10 keV to 20 keV or 40 keV, the epithermal neutron flux is more than 109 cm−2 s−1 at a proton energy of 3 MeV for Li target and 8 MeV for Be target.

Electric field design of cold cathode Penning ion source for miniaturization neutron tube

It is of great significance to study the optimum electric field strength of cold cathode Penning ion source (PIG), which is used on high yield neutron tube. The electric field in the ion source depends largely on the relationship between the voltage of anode and the thickness of anticathode. In the case of constant magnetic strength, the discharge modes of PIG can be distinguished by the relation between electric potential and pressure. Based on this relationship, we re-designed and optimized the PIG, and got the expected discharge mode. COMSOL was used to simulate the model, and a better extracted beam intensity (10.31μA) and a higher proton fraction (24 %) were obtained at an extracted voltage of −80 kV. It is estimated that the yield is higher under −80 kV than the previous design, which can reach 6×108n/s. This study can provide reference for the development of high yield neutron tube ion sources.

Radioactivity study of EAST structural materials based on neutron activation system

The plasma operation in the Experimental Advanced Superconducting Tokamak (EAST) would generate 2.45 MeV neutrons and a small fraction of 14 MeV neutrons. Neutrons react with structural materials, leading to the production of radioactive isotopes. The identification of nuclides and the determination of radioactive activity are of great significance for radiation protection and fusion material applications. In this study, structural materials including stainless steel 316L(SS316L), stainless steel 304L(SS304L), aluminum(Al), tungsten(W) and copper(Cu) were transported to the activation terminal for fusion neutron irradiation. The gamma ray spectra were measured using high-purity germanium (HPGe) detector, and the main radioactive nuclides of SS316L and SS304L are 56Mn, while the main radioactive nuclide of Al is 28Al. For Cu and W, the main radionuclides resulting from neutron reactions are 66Cu and 187W, respectively. The uncertainty of the radioactivity is 2.27∼7.94 % for W, 1.95∼5.84 % for SS316L, 2.10∼6.07 % for SS304L, 0.68∼1.69 % for Al and 2.59∼6.75 % for Cu. The specific activity of radionuclides is proportional to neutron yield, so the material activation is the main source of radiation dose during irradiation with high-yield neutron. The neutron activation coefficients of SS316L, SS304L, Al, Cu and W are 2.74538 × 10−14, 3.6411 × 10−14, 2.50711 × 10−12, 1.76506 × 10−12 and 3.51725 × 10−14, respectively. The coefficient represents the radioactive activity generated by each fusion neutron. In addition, the predicted activity was calculated using FISPACT program and compared with experiments.

Physics design of 14 MeV neutron generator facility at the Institute for Plasma Research

A high energy and high yield neutron source is a prime requirement for technological studies related to fusion reactor development. It provides a high-energy neutron environment for small-scale fusion reactor components research and testing such as tritium breeding, shielding, plasma-facing materials, reaction cross-section data study for fusion materials, etc. Along with ITER participation, the Institute of Plasma Research, India is developing an accelerator-based 14 MeV neutron source with a yield of 1012 n s−1. The design of the source is based on the deuterium–tritium fusion reaction. The deuterium beam is accelerated and delivered to the tritium target to generate 14 MeV neutrons. The deuterium beam energy and tritium availability in the tritium target are the base parameters of the accelerator-based neutron source design. The paper gives the physics design of the neutron generator facility of the Institute for Plasma Research. It covers the requirements, design basis, and physics parameters of the neutron generator. As per the analytical results generator can produce more than 1 × 1012 n s−1 with a 110 keV D+ ion beam of 10 mA and a minimum 5 Ci tritium target. However, the detailed simulation with the more realistic conditions of deuteron ion interaction with the tritium titanium target shows that the desired results cannot be achieved with 110 keV. The safe limit of the ion energy should be 300 keV as per the simulation. At 300 keV ion energy and 20 mA current, it reaches 1.6 × 1012 n s−1. Moreover, it was found that to ensure sufficiently long operation time a tritium target of more than 20 Ci should be used. The scope of the neutron source is not limited to the fusion reactor research studies, it is extended to other areas such as medical radioisotopes research, semiconductor devices irradiations, and many more.

The magnetic field design of a solenoid for the cold-cathode Penning ion source of a miniature neutron tube

A high-yield and beam-stable neutron tube can be applied in many fields. It is of great significance to the optimal external magnetic field intensity of the cold-cathode Penning ion source (PIS) and precisely controls the movement of deuterium (D), tritium (T) ions and electrons in the source of the neutron tubes. A cold-cathode PIS is designed based on the solenoidal magnetic field to obtain better uniformity of the magnetic field and higher yield of the neutron tube. The degree of magnetic field uniformity among the magnetic block, double magnetic rings and solenoidal ion sources is compared using finite element simulation methods. Using drift diffusion approximation and a magnetic field coupling method, the plasma distribution of hydrogen and the relationship between plasma density and magnetic field intensity at 0.06 Pa pressure and a solenoid magnetic field are obtained. The results show that the solenoidal ion source has the most uniform magnetic field distribution. The optimum magnetic field strength of about 0.1 T is obtained in the ion source at an excitation voltage of 1 V. The maximum average number density of monatomic hydrogen ions (H+) is 1 × 108 m−3, and an ion-beam current of about 14.51 μA is formed under the −5000 V extraction field. The study of the solenoidal magnetic field contributes to the understanding of the particle dynamics within the PIS and provides a reference for the further improvement of the source performance of the neutron tube in the future.

Ultrashort pulsed neutron source driven by two counter-propagating laser pulses interacting with ultra-thin foil

Neutron production via D(d, n)<sup>3</sup>He nuclear reaction during the interaction of two counter-propagating circularly polarized laser pulses with ultra-thin deuterium target is investigated by particle-in-cell simulation and Monte Carlo method. It is found that the rotation direction and initial relative phase difference of laser electric field vector have important effects on deuterium foil compression and neutron characteristics. The reason is attributed to the net light pressure and the difference in transverse instability development. The highest neutron yield can be obtained by choosing two laser pulses with a relative phase difference of 0 and the same rotation direction of the electric field vector. When the relative phase difference is 0.5π or 1.5π and the rotation direction of electric field vector is different, the neutrons have a directional spatial distribution and the neutron yield only slightly decreases. For left-handed circularly polarized laser pulse and right-handed circularly polarized laser pulse, each with an intensity of 1.23 × 10<sup>21</sup> W/cm<sup>2</sup>, a pulse width of 33 fs and a relative phase difference of 0.5π, it is possible to produce a pulsed neutron source with a yield of 8.5 × 10<sup>4</sup> n, production rate of 1.2 × 10<sup>19</sup> n/s, pulse width of 23 fs and good forward direction as well as tunable spatial distribution. Comparing with photonuclear neutron source and beam target neutron source driven by ultraintense laser pulses, the duration of neutron source in our scheme decreases significantly, thereby possessing many potential applications such as neutron nuclear data measurement. Our scheme offers a possible method to obtain a compact neutron source with short pulse width, high production rate and good forward direction.

A new dedicated signal processing system for gamma-ray spectrometers in high power deuterium-tritium plasma scenarios in tokamaks.

The most performant deuterium-tritium (DT) plasma discharges realized by the Joint European Torus (JET) tokamak in the recent DT campaign have produced neutron yields on the order of 1018 n/s. At such high neutron yields, gamma-ray spectroscopy measurements with scintillators are challenging as events from the neutron-induced background often dominate over the signal, leading to a significant fraction of pileup events and instability of the photodetector gain along with the consequent degradation of the reconstructed spectrum. Here, we describe the solutions adopted for the tangential lanthanum bromide spectrometer installed at JET. A data acquisition system with free streaming mode digitization capabilities for the entire duration of the discharge has been used to solve dead-time related issues and a data reconstruction code with pileup recovery and photodetector gain drift restoration has been implemented for off-line analysis of the data. This work focuses on the acquired data storage and parsing, with a detailed explanation of the pileup recovery and gain drift restoration algorithms.

Characterization of the hardened single line of sight camera at the National Ignition Facility.

The hardened single line of sight camera has been recently characterized in preparation for its deployment on the National Ignition Facility. The latest creation based on the pulse-dilation technology leads to many new features and improvements over the previous-generation cameras to provide better quality measurements of inertial confinement fusion experiments, including during high neutron yield implosions. Here, we present the characterization data that illustrate the main performance features of this instrument, such as extended dynamic range and adjustable internal magnification, leading to improved spatial resolution.

Pulsed neutron source from interaction of relativistic laser pulse with micro-structure assisted pitcher–catcher target

Neutron detection technology is a powerful method for characterizing fundamental and industrial materials, such as magnetic materials, nanomaterials, polymers, and biological substances. A compact pulsed neutron source is desired, as the traditional way of producing high-yield neutrons depends mainly on large-scale accelerators or fission reactors. Here, we propose a scheme to generate a high-yield pulsed neutron source by using a 100 fs relativistic laser pulse interacting with a micro-structure assisted pitcher–catcher target. Three-dimensional particle-in-cell and Monte Carlo hybrid simulations demonstrate that an energetic deuterium ion beam with a cutoff energy of 75 MeV and a 7.23 laser-to-deuterium energy conversion efficiency can be obtained with a laser pulse of intensity W cm−2, duration 330 fs, power 34 TW and energy 6.7 J. When they strike the following lithium fluoride converter of thickness 2 cm, a large number of neutrons are thus produced via a 7Li(d,n) nuclear reaction. The neutron yield is up to 109 and its pulse duration is as short as 20 ps. This scheme could be realized in laboratories with current hundreds-of-terawatt or multi-petawatt laser facilities.

Neutron transmission imaging with a portable D-T neutron generator

A portable fast-neutron imaging system is being developed to provide complementary information to field X-ray imaging. Applications include inspection of vehicles and infrastructure for corrosion, measurement of material levels in containers, and inspection of munitions and suspicious packages. While fast-neuron imaging generally provides lower imaging resolution compared to X-rays, fast-neutron interaction cross-sections have a weak dependence on material Z. This enables imaging of low-Z materials inside high-Z materials. Here, we discuss the limitations and current improvements in fast-neuron imaging. Limitations in portable fast-neutron imaging systems include low D-T neutron generator output, low light production in ZnS(Cu) imaging scintillators, low resolution due to scintillator thickness and D-T spot size, and digital-panel darknoise that varies in time and position and that can be 100× larger than the neutron signal. We have made improvements in these areas through development of a segmented high light yield scintillator, panel noise mitigation techniques, and testing of new high-output, small spot size D-T neutron generators. The segmented high light yield fast-neutron scintillator demonstrated 5× increase in light compared to ZnS(Cu). An additional 2× improvement in signal-to-noise was demonstrated with panel-noise mitigation techniques. Our MCNP calculations also show good agreement with neutron imaging results We have demonstrated improvements in fast-neutron imaging through development of a segmented high light yield neutron scintillator, mitigation of digital panel noise, and preliminary testing with new high-output, small spot size D-T neutron generators. We have also demonstrated good results modeling fast-neutron images and scatter effects using MCNP.

The influence of target film material and coating on neutron yield and sputtering yield

The long-life, high yield deuterium-deuterium (D-D) neutron tube has become one of the research hotspots. Here, deuterated polyethylene target, heavy water target and titanium target were investigated by Stopping and Range of Ions in Matter (SRIM). The calculation showed that the deuterated polyethylene target, which was a potential target material, had the highest yield at an incident energy of 120 keV. Further, considering the unfavorable factors such as impurity ions and high temperature, the coating was used to protect the target material. Diamond, boron carbide, boron nitride, silicon carbide, and aluminum carbide were selected. The simulation results showed that the diamond composite deuterated polyethylene film had the best sputtering resistance, and the aluminum nitride composite heavy water target film had the lowest sputtering yield. The two coating materials shield the target, reduced the energy loss of incident ions, and provided a new method for the research of high yield and long life neutron tube.

АНАЛИЗ ПОЯВЛЕНИЯ МИКРОЯДЕР В ЭРИТРОЦИТАХ И АКТИВНОСТИ ПРОЛИФЕРАЦИИ КЛЕТОК КОСТНОГО МОЗГА ПОСЛЕ ПРОЛОНГИРОВАННОГО ОБЛУЧЕНИЯ МЫШЕЙ БЫСТРЫМИ НЕЙТРОНАМИ В НИЗКИХ ДОЗАХ

Purpose: To analyze the level of cytogenetic damage and the activity of bone marrow cells proliferation in C57BL/6 mice after prolonged fast neutrons low dose irradiation at 10–500 mGy. 
 Material and methods: Male C57BL/6 mice at the age of 7–8 and 16 weeks were used in the experiments. Irradiation was carried out on an OR-M installation in the field of fast neutrons and gamma quanta using five Pu(α,n)Be radionuclide sources with a high fast neutron yield at a dose rate of 2.13 mGy/h. The frequency of polychromatophilic (PCE) and normochromic (NCE) erythrocytes with micronuclei (MN) and the ratio of PCE and NCE were analyzed using light microscopy after cytochemical staining of the bone marrow cells of control and irradiated mice. The proliferation activity of bone marrow cells was determined by the number of Ki-67+-cells. The parameters of the cell cycle and the level of apoptosis were studied after DNA staining with DAPI using flow cytometry. Statistical processing of the results was carried out according to the Student’s method using the computer program Origin.
 Results: It was found that prolonged irradiation of mice with fast neutrons at a low dose rate (2.13 mGy/h) at doses from 10 to 500 mGy after 24 h led to statistically significant increase in the frequency of PCE with MN at all studied doses. No dose dependence of this parameter was observed in the studied range. The increase in the frequency of PCE with MN at a dose of 500 mGy was prolonged and persisted for at least 72 h. A significant increase in the frequency of NCE with MN 24 h after irradiation was found only at a dose of 500 mGy, which persisted up to 48 h. At this dose, there was also a decrease in the number of nucleated cells in the bone marrow 24 – 72 h after exposure, a decrease in the number of Ki-67+-cells 24 h after irradiation of mice, a block of the cell cycle in the G2/M phase, and a decrease of cells in the G0/G1 phase, but after 48 h, there were no disturbances in the cell cycle. 
 Conclusion: It has been shown that after a single total prolonged irradiation of mice at low doses (10–500 mGy), when analyzing the frequency of PCE with MN, cytogenetic damage is recorded in the bone marrow, which indicates the genetic danger of exposure to even such low levels of fast neutron irradiation. A decrease in Ki67+ cells and cell cycle arrest at the G2/M phase were found only after irradiation of mice at a dose of 500 mGy and only 24 h after exposure, while the number of nucleated cells in the bone marrow at this dose was reduced, at least to 72 h.

MCNP Dose evaluation around D-D and D-T neutron generators and shielding design

In recent years, neutron generators have extensively used as neutron sources and vast efforts have been made to develop high yield neutron generators, so the aspect of radiation shielding during the utilization of neutron generators is unavoidable. In this study, dose distribution around the D-D and D-T neutron generator was calculated by using Monte Carlo simulation. Then an effective shielding was designed and different materials with different thicknesses at distinction distances from the generator were examined. Finally, the organs dose was acquired by using ICRP 110 male phantom with and without shielding at different body position relative to generator tube. The results show that the dose induced by the D-T generator is 500 times greater than the D-D generator and the neutron dose is about 100 times more than gamma dose. The dose is reduced about 20 times by increasing the distance up to 5 m, however, the presence of Borated-Polyethylene with a thickness of 60 cm as most effective shielding reduces the whole-body dose up to 60%. Besides, inserting a layer of Pb to Borated-Polyethylene is effective to attenuate the gamma-rays. The received dose to organs depends on the body direction relative to tube head of neutron generator. In general, the best position of tube placement relative to shielding wall are 90 and 180 degrees for the D-T and D-D generators, respectively, distance between the generator tube head and the phantom is suggested 200 cm and it is recommended that for most dose reduction, the operator stands in the anterior-posterior position relative to generator tube.