Deuterium-tritium Neutron Research Articles

Nonlinear light-output calibration of the oxygenated xylene scintillators used in OMEGA neutron time-of-flight spectrometers.

Neutron time-of-flight (nTOF) spectrometers are essential instruments for measuring and evaluating the performance of inertial confinement fusion implosions. The neutron spectrometers utilized for the OMEGA laser include two liquid-based scintillators, each consisting of a large volume filled with xylene that is coupled to four photomultiplier tubes. Analysis of the signal from these detectors requires detailed knowledge of the scintillator's light output, which is needed to fit the nTOF spectrum, from which the neutron energy spectrum is informed. The light output is nonlinearly proportional to the neutron energy, which, in turn, affects the interpretation of the neutron energy spectrum from a TOF signal. A recent campaign on OMEGA was performed to calibrate the xylene detectors and infer the shape of the light-output curve. The campaign utilized materials with increasing Z placed in the OMEGA target chamber to initiate scattering events with the 14MeV fusion neutrons. This process leads to the production of backscatter neutrons of varying energies that appear as peaks in the nTOF data. Simulations using a neutron transport code were combined with the measured deuterium-tritium neutron yields to calculate the expected backscattered neutron yields from the well-known scattering cross sections of each material. The neutron-energy dependent light output of the scintillator inferred from the experiment is compared to the light-output curve simulated with a neutron transport code for the following neutron energies: 1.5, 2.5, 6, and 14MeV.

Diagnosing up-scattered deuterium-tritium fusion neutrons produced in burning plasmas at the National Ignition Facility (invited).

In the push to higher performance fusion plasmas, two critical quantities to diagnose are α-heat deposition that can improve and impurities mixed into the plasma that can limit performance. In high-density, highly collisional inertial confinement fusion burning plasmas, there is a significant probability that deuterium-tritium (DT) fusion products, 14.1MeV neutrons and 3.5MeV α-particles, will collide with and deposit energy onto ("up-scatter") surrounding deuterium and tritium fuel ions. These up-scattered D and T ions can then undergo fusion while in-flight and produce an up-scattered neutron (15-30MeV). These reaction-in-flight (RIF) neutrons can then be uniquely identified in the measured neutron energy spectrum. The magnitude, shape, and relative size of this spectral feature can inform models of stopping-power in the DT plasma and hence is directly proportional to α-heat deposition. In addition, the RIF spectrum can be related to mix into the burning fuel, particularly relevant for high-Z shell and other emerging National Ignition Facility platforms. The neutron time-of-flight diagnostic upgrades needed to obtain this small signal, ∼10-5 times the primary DT neutron peak, will be discussed. Results from several gain > 1 implosions will be shown and compared to previous RIF spectra. Finally, comparisons of experimental data to a simplified computational model will be made.

Simulated neutron dose for two cell culture configurations

One of the major hurdles for long duration interplanetary space missions is space radiation-induced carcinogenesis. The need for early identification of relevant biomarkers of elevated space radiation risk is crucial for ensuring the health of astronauts after returning from long-duration exploration missions. The longer the duration of the mission, the greater the risk of space-radiation induced carcinogenesis. It is vital to have an accurate understanding of the absorbed dose to sensitive cells comprising the liver and colon cell models for identifying biomarkers that should be further investigated in more human-like systems exposed to space-like radiation. The specific liver and colon cell lines were chosen to build on previous experience with them and to establish a proof of concept for investigation of individual variability impact on biomarkers. Monte Carlo N-Particle transport code 6.2 was used in the simulation of two-dimensional and three-dimensional cell cultures exposed to deuterium–tritium fusion neutrons. Fast neutrons were chosen as most space neutron dose is delivered by neutrons with energy greater than 100MeV. Deuterium–tritium fusion neutrons posses the greatest possible neutron energy outside of large accelerator facilities. Deuterium–tritium neutron-induced charged particles closely resemble the high-LET portion of the GCR spectrum. The absorbed dose was recorded over the cell layer (target volume) using an F6 tally in MCNP. From foil activation analysis, the experimental source strength of 1.2 × 108 neutrons s−1 was used to determine the absorbed dose rate to the cells, which ranged from 69 to 73 mGy h−1. This simulated dose rate will be used to determine the total dose received by the cells in experiments conducted to identify biomarkers related to increased risk of carcinogenesis from fast neutron exposure.

Optimization and experimental validation of deuterium-tritium neutron monoenergetics for diamond detector testing

Diamond detectors are excellent tools for measuring deuterium-tritium (D-T) neutron energy spectra. It has great potential for application in the field of magnetic confinement fusion. However, the current experimental values of the energy resolution of diamond detectors are much lower than the theoretical estimates. The purpose of this study is to find a method for experimental optimization. For this purpose, the D-T neutron energy distribution of the accelerator is theoretically analyzed and simulated. A method to increase the deuteron energy to obtain a better monoenergetic neutron source is proposed. This strategy was validated experimentally. The energy resolution of diamond for D-T neutrons achieves 1.21%, a Full Width at Half Maximum of 93 keV at 7.65 MeV. It is the highest resolution recorded to date. These results demonstrate the feasibility of the optimization method.

Response of a high-pressure 4He scintillation detector to nuclear recoils up to 9 MeV

Helium-4-based scintillation detector technology is emerging as a strong alternative to pulse-shape discrimination-capable organic scintillators for fast neutron detection and spectroscopy, particularly in extreme gamma-ray environments. The 4He detector is intrinsically insensitive to gamma radiation, as it has a relatively low cross-section for gamma-ray interactions, and the stopping power of electrons in the 4He medium is low compared to that of 4He recoil nuclei. Consequently, gamma rays can be discriminated by simple energy deposition thresholding instead of the more complex pulse shape analysis. The energy resolution of 4He scintillation detectors has not yet been well-characterized over a broad range of energy depositions, which limits the ability to deconvolve the source spectra. In this work, an experiment was performed to characterize the response of an Arktis S670 4He detector to nuclear recoils up to 9 MeV. The 4He detector was positioned in the center of a semicircular array of organic scintillation detectors operated in coincidence. Deuterium–deuterium and deuterium–tritium neutron generators provided monoenergetic neutrons, yielding geometrically constrained nuclear recoils ranging from 0.0925 to 8.87 MeV. The detector response provides evidence for scintillation linearity beyond the previously reported energy range. The measured response was used to develop an energy resolution function applicable to this energy range for use in high-fidelity detector simulations needed by future applications.

Novel deuterium-tritium (DT) neutron activation diagnostics for increased neutron yield operations at Sandia's Z-facility (invited).

As inertial confinement fusion experiments at Sandia National Laboratories' Z-facility begin using deuterium-tritium (DT) fuels, the expected DT-neutron yields are expected to increase many orders of magnitude. This brings about the need for diagnostics that have a significantly higher dynamic range of operation and ability to obtain yields where there will be significant radiation holds. Using the current activation diagnostic infrastructure at the Z-facility, a pair of activation diagnostics are being developed. The first is fluorine activation, which will be used in a similar manner to current copper activation diagnostics that are fielded in a rabbit for quick removal from the experiments without opening the vacuum experiment containment. The second is a prompt DT-neutron detector that exploits the activation of oxygen with similar operation characteristics as the current Be-probes fielded at Z. We will discuss the design of this pair of instruments, along with calibration work performed at Sandia's Ion Beam Laboratory in support of the development and anticipated instrument performance.

Radiation Techniques for Tracking the Progress of the Hydrometallurgical Leaching Process: A Case Study of Mn and Zn

With advancements in hardware and software, non-destructive radiometric analytical methods have become popular in a wide range of applications. A typical case is the study of the leaching process of metals from mineral ores and mine tailings. The objective of the current study was to develop a radiometric method based on neutron activation analysis (NAA), in particular, delayed gamma neutron activation analysis (DGNAA), to monitor the process of Mn and Zn leaching from Ti ore, Cu mine tailings, and Zn-Pb mine tailings. The DGNAA method was performed using a neutron source: a deuterium-tritium (D-T) neutron generator for Mn and a MARIA research nuclear reactor for Zn. Laboratory-scale Mn leaching from Ti ores, Cu tailings, and Zn-Pb tailings was investigated using delayed gamma-rays of 56Mn (half-life of 2.6 h). The dissolution efficiencies of Mn were found to increase with interaction time and HCl concentration (1 to 5 M) and to vary with the leaching temperature (22.5 to 110 °C). Such results were found to agree with those obtained by total reflection X-ray fluorescence (TXRF) spectrometry for the same samples. 65Zn (half-life of 244 days) was chosen to investigate real-time/online leaching of Zn in Ti ore, Cu tailings, and Zn-Pb tailings. During online monitoring, Zn recovery was also reported to increase with increased leaching time. After approximately 300 min of leaching, 80%, 79%, and 53% recovery of Zn in Zn-Pb tailings, Ti ore, and Cu tailings, respectively, were reported. Theoretically, developed mathematical prediction models for 65Zn radiotracer analysis showed that the spherical diffusion model requires much less time to attain saturation compared to the linear diffusion model. The results of NAA for Zn were compared with those obtained by handheld X-ray fluorescence (handheld-XRF) and TXRF analysis. The analyzed samples encompassed leached Ti ore, Cu tailings, and Zn-Pb tailings which were subjected to different conditions of leaching time, temperature, and HCl concentrations. The XRF analysis confirmed that the leaching efficiencies of Zn rise with the increase in leaching time and HCl concentration and fluctuate with leaching temperature. The developed approach is important and can be applied in laboratories and industrial setups for online monitoring of the recovery of any element whose isotopes can be activated using neutrons. The efficiency of the metal-recovery process has a direct impact on the normal operation and economic advantages of hydrometallurgy.

Operation of a 250μm-thick SiC detector with DT neutrons at high temperatures

The Silicon Carbide detector (SiC) is an object of research as an alternative to diamond detectors for fast neutron detection and spectrometry where harsh environments are an issue, like in Tokamaks. Since future breeding blankets mock-ups will feature temperatures up to 550 °C, diamond detectors were characterized in the past, finding limitations in their functionality at high temperatures. This paper expands on the previous work by proving the detection of fast neutrons with good detection parameters of a 250 μm-thick 4 H-SiC detector prototype at temperatures up to 250 °C, highlighting the detector's excellent resilience to temperature. The experiment is conducted with instrumentation similar to the one used in the past with diamond detectors, using as source of irradiation the Frascati Neutron Generator (FNG) in ENEA, which is accelerator driven neutron source based on deuterium-tritium (DT) fusion reaction.

OpenMC Interpretation of FNS SINBAD Shielding Benchmark Experiments

The Fusion Neutron Source (FNS) clean benchmark experiments on tungsten, vanadium, and beryllium assemblies from the SINBAD (Shielding Integral Benchmark Archive and Database) are analyzed to experimentally validate OpenMC (version 0.14.1-dev) fusion neutronics capabilities. The assemblies were irradiated with a 14-MeV deuterium-tritium neutron source. Neutron spectra, photon spectra, reaction rates, gamma heating rates (GHRs), and tritium production rates (TPRs) are compared to measured data in the experimental assemblies and MCNP-6.2 results. In the tungsten case, slight overestimations of the experimental data were observed in the neutron spectra, and the photon spectra agreed well with the experiments. Most of the GHRs agreed with the measured data within the range of experimental uncertainty in the tungsten and vanadium assemblies. In the vanadium assembly, the calculated neutron spectra underestimated the experiments in the low energy region while the photon spectra were well calculated when compared to experiments. The most noticeable discrepancies with experimental data in the gamma heating were observed at detector positions closest to the source. For the reaction rates, notable discrepancies with experimental data were seen at the front and rear of the assemblies. Compared to experiments, the OpenMC neutron spectra were well predicted in the beryllium assembly, whereas the calculated fission reaction rate and TPRs overestimated the experiments, an observation similar to that which has been reported by other authors. The average, overall calculation-to-experiment ratio (C/E) over nine TPR and seven GHR measurements were 1.03 ± 0.20 and 0.95 ± 0.14, respectively. In the case of verification, the OpenMC results of the benchmark calculations indicated comparable accuracy to MCNP-6.2. In general, the validation exercise showed that OpenMC can be used to analyze the fusion neutronics shielding benchmark problems.

Characterisation of neutron fields at the n-lab, a fast neutron facility at the University of Cape Town

The n-lab is a fast neutron facility based in the Department of Physics, University of Cape Town, offering collimated neutron beams produced by an MP 320 deuterium–tritium sealed tube neutron generator, and a 220 GBq americium–beryllium radioisotopic source. Characterisations of the spatial and energy distributions of the fast neutron beams were performed using an EJ-301 organic liquid scintillator detector and digital data acquisition system. Neutron energy spectra were obtained through unfolding analyses with MAXED, and a Monte Carlo approach to the propagation of uncertainties was implemented. Measurements of fluence and neutron emission rates were determined through the neutron activation of copper foils and subsequently used to validate the scaling of the unfolded neutron energy spectra.

Design and simulated performance of a high-resolution magnetic proton recoil spectrometer for deuterium–tritium neutrons

Neutron emission energy and time spectra of deuterium–tritium plasma contain abundant information about the fusion yield, ion temperature, and nuclear burn history. Magnetic recoil spectrometers have been employed in measuring time-integrated energy spectra in inertial confinement fusion (ICF) experiments. As the fusion yield increases, time-resolved neutron diagnostics is urgently desired for understanding complex fusion physics. Herein, we design a compact magnetic proton recoil (MPR) spectrometer based on a 90 degree permanent magnet. The entrance and exit of the magnet are second-order shapes to eliminate two major second-order aberrations. The energy resolution and time resolution for 14MeV neutrons achieve 97keV and 1.6ns, respectively, under the detection efficiency of 7.4×10−8cm2. A quadratic relation is found between detection efficiency and energy resolution when configurations are optimal. Monte Carlo simulations are accomplished to study the performance in measuring time-integrated energy spectra, emission time spectra, and time-resolved energy spectra, demonstrating the potential of high-resolution MPR spectrometers in future high-yield ICF experiments.

Evidence of non-Maxwellian ion velocity distributions in spherical shock-driven implosions.

The ion velocity distribution functions of thermonuclear plasmas generated by spherical laser direct drive implosions are studied using deuterium-tritium (DT) and deuterium-deuterium (DD) fusion neutron energy spectrum measurements. A hydrodynamic Maxwellian plasma model accurately describes measurements made from lower temperature (<10keV), hydrodynamiclike plasmas, but is insufficient to describe measurements made from higher temperature more kineticlike plasmas. The high temperature measurements are more consistent with Vlasov-Fokker-Planck (VFP) simulation results which predict the presence of a bimodal plasma ion velocity distribution near peak neutron production. These measurements provide direct experimental evidence of non-Maxwellian ion velocity distributions in spherical shock driven implosions and provide useful data for benchmarking kinetic VFP simulations.

Determination of boron content in Tinkal ore samples with Geant4 using MCLLS method

A new Monte Carlo-Library Least Squares (MCLLS) method was developed for the neutron inelastic-scattering and thermal-capture analysis (NITA) techniques in real-time online analysis systems to determine boron content in Tinkal ore samples. The Geant4 toolkit was used to simulate the Prompt Gamma-ray Neutron Activation Analysis (PGNAA) system for the NITA technique. Tinkal ore samples were bombarded with 14 MeV neutron beams in order to represent neutrons released from the Deuterium-Tritium (DT) neutron generator. After the simulation of shooting Tinkal ore samples with 14 MeV neutron beams was completed, histograms were obtained from Geant4. Once the histograms are broadened according to the shape calibration of the BGO scintillation detector, the spectrum (broadened histograms) are ready to be used in the MCLLS process. A single element library was created by broadened histograms of the nine Tinkal ore samples. In addition, set #10 was selected to be an unknown sample and was used for validating MCLLS. After analysing the unknown sample spectrum, boron content was estimated by the single element library created by the MCLLS method. The estimated value of the boron content by MCLLS was consistent with the laboratory result.

Neutron collimator optimization for 14.1 MeV DT neutrons using Monte Carlo and Genetic algorithms

The fast neutrons generated by Deuterium-Tritium (DT) fusion reaction have been widely applied in prompt gamma ray neutron activation analysis measurements. In this study, a multi-layer neutron collimator for DT neutron generator was developed. Genetic algorithm combined with Monte Carlo simulation was used to design a collimator made of iron, lead, graphite, and borated polyethene. Copper foil activations were conducted to determine the fast neutron flux ratios between the beam port and its nearby area and agreed well with those predicted by the simulations. The results demonstrated that a narrower beam was obtained. The fast neutron beam flux was 568 ± 14 s−1 cm−2. The neutron flux ratio of the collimator was improved by a factor of 2.36, which could provide a better neutron beam.

Study on properties of boron bearing blast furnace slag composite as shielding material for D-T neutron generator

Recovery of boron bearing blast furnace slag (B3FS) as a highly cost-effective radiation shielding material can reduce radiation hazards and protect the environment. Four different shielding configurations based on the Monte Carlo Particle Transport Program (MCNP) has been established to calculate the percentage for neutron shielding, radiation dose and neutron flux to evaluate the shielding properties of B3FS composites for Deuterium-Tritium (D-T) neutron. The properties of the B3FS composite are compared with common commercial shielding materials, include boron carbide (B4C), barite concrete, boron containing polyethylene (BPE), high-density polyethylene (HDPE), ordinary concrete, Portland concrete and lead. The results indicated that the B3FS composite has a high neutron shielding property, which is close to that of B4C, BPE, HDPE and some kinds of concrete under shielding plate configuration. The neutron shielding properties of the composite are similar to those of BPE and HDPE, and are better than other shielding materials under shielding material wrapping device configuration. Based on the results, B3FS composite can be used as a new type of neutron shielding material with excellent neutron shielding performance and environmental protection.

Neutronic simulations of in-vessel neutron calibrations for ITER neutron diagnostics by using simplified ITER model

The absolute calibration of the detection efficiency for the measurement of the total neutron yield in the whole plasma is one of the most important issues in neutron diagnostics. In ITER, a compact deuterium-tritium (D-T) neutron generator will be used as a neutron source for the neutron calibration prior to the D-T plasma experiments, where the neutron source will be moved in the vacuum vessel. Neutronic simulations of neutron diagnostic calibrations have been carried out for the strategy and scheduling of the calibration experiments. We have established a simplified 360° ITER model for the simulation, which includes neutron flux monitors (NFMs) in an equatorial port (EQ), micro fission chambers, diverter neutron flux monitors, and a neutron activation system. At first, angular neutron spectra of the compact D-T neutron generator have been evaluated by MCNP calculations. We evaluate the discrepancies among the detection efficiencies to be obtained by the neutron calibration experiment, those by idealistic D-T ring source, and actual detection efficiencies for the plasma neutron source. Point efficiency measurement using the compact D-T neutron generator facing NFM in EQ#1 is an effective method for the NFM calibration.

Neutron imaging of inertial confinement fusion implosions.

We review experimental neutron imaging of inertial confinement fusion sources, including the neutron imaging systems that have been used in our measurements at the National Ignition Facility. These systems allow measurements with 10 µm resolution for fusion deuterium-deuterium and deuterium-tritium neutron sources with mean radius up to 400 µm, including measurements of neutrons scattered to lower energy in the remaining cold fuel. These measurements are critical for understanding the fusion burn volume and the three-dimensional effects that can reduce the neutron yields.

Life prediction of neutron tube based on regression analysis and Wiener process

Life prediction is a complicated problem to be solved for the neutron tube which is widely used and plays an important role in many industries. In this paper, research on life prediction using regression analysis and Wiener process was carried out in order to evaluate the service life of the neutron tube quickly. Failure modes of the prefabricated target sealed deuterium-tritium neutron tube were firstly studied. Life tests of the neutron tube were conducted during laboratory at room temperature for about 55 hours. Regression analysis was performed on the experimental data obtained by the life tests of the neutron tube. The service life of the neutron tube was predicted using the regression function and the uncertainty analysis was performed. In order to take the stochasticity of the performance degradation and the yield measurement of the neutron tube into consideration, the Wiener process in which the Kalman filter and the maximum likelihood estimation methods were applied for the parameter solving process was adopted to assess the service life of the neutron tube. A program of the Wiener process was developed and the residual life distribution of the neutron tube was obtained. The predicted life which consisted of the expectation of the residual life and the time of life tests was consistent with that using the regression function. Finally, the scope of applications of the methods and the results in this paper were discussed. The general process in this paper may be used to estimate the service life of the neutron tube in the early stage of development.

Alpha heating of indirect-drive layered implosions on the National Ignition Facility.

In order to understand how close current layered implosions in indirect-drive inertial confinement fusion are to ignition, it is necessary to measure the level of alpha heating present. To this end, pairs of experiments were performed that consisted of a low-yield tritium-hydrogen-deuterium (THD) layered implosion and a high-yield deuterium-tritium (DT) layered implosion to validate experimentally current simulation-based methods of determining yield amplification. The THD capsules were designed to reduce simultaneously DT neutron yield (alpha heating) and maintain hydrodynamic similarity with the higher yield DT capsules. The ratio of the yields measured in these experiments then allowed the alpha heating level of the DT layered implosions to be determined. The level of alpha heating inferred is consistent with fits to simulations expressed in terms of experimentally measurable quantities and enables us to infer the level of alpha heating in recent high-performing implosions.

Validation of S(α, β) thermal neutron scattering libraries using pulsed-neutron die-away experiments

Validation of the accuracy of S(α, β) thermal scattering law (TSL) evaluations for moderator materials is an important task for the development of high-performance nuclear engineering systems. Many recent thermal neutron scattering evaluations have had limited experimental validation. Like other nuclear data, validation of TSL libraries has historically been by integral criticality benchmarks. While sufficient for general study, these benchmarks often have limited sensitivity to the tested TSLs, and compounding uncertainties from other nuclear data can make validation ambiguous. In some cases, no criticality benchmarks exist that are sensitive to the TSLs of interest. With the development of high-performance next-generation thermal nuclear reactors, alternative validation of applicable TSLs is of high importance. By performing thermal neutron measurements via pulsed-neutron die-away (PNDA) experiments, along with parallel simulations, the integral performance of various TSL evaluations can be compared to measured experimental data. An experimental testbed using a D-T neutron generator, moderator sample, and thermal neutron detector was assembled at Rensselaer Polytechnic Institute. A Thermo Scientific D211 Deuterium-Tritium Neutron generator is used to generate 10 μs neutron pulses. Various targets of different sizes and geometries are used to moderate the neutrons. Multiple detector types and configurations were tested to optimize the experiment. The room-temperature polyethylene TSL evaluation is well vetted and is similar across different evaluations. This makes it an ideal evaluation to compare with the experimental results.