Deuterium-tritium Neutron Source Research Articles

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.

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.

Time-resolved secondary triton burnup 14 MeV neutron measurement by a new scintillating fiber detector in middle total neutron emission ranges in deuterium large helical device plasma experiments

A middle-sensitivity scintillating fiber detector (hereafter middle Sci-Fi detector) that works at a deuterium-tritium neutron flux of ~105-107 cm−2s−1 was utilized to measure secondary deuterium-tritium neutron emission rates with high temporal resolution at a total neutron emission rate of 1013 to 1015 n/s, where strong magnetohydrodynamic (MHD) instabilities were observed in the large helical device deuterium plasma experiments. The gain and angular characteristics of the middle Sci-Fi detector were evaluated in an accelerator-based deuterium-tritium neutron source in the intense 14 MeV neutron source facility at Osaka University. Observation of 1 MeV triton transport due to MHD instability was performed by a middle Sci-Fi detector whose deuterium-tritium neutron counting rate was approximately 20 times higher than that of the conventional Sci-Fi detector. Fusion-born triton transport due to energetic-particle-driven MHD instability was observed using the middle Sci-Fi detector due to its high detection efficiency and high discrimination ability of deuterium-tritium neutrons from the sea of deuterium-deuterium neutrons.

Bulk material interrogation experimental results and validation with Geant4 for replacement of dangerous radiological sources in oil-well logging industries

The Kansas State University Materials Interrogation (KSUMI) test facility was set up to enable bulk-material irradiation experiments that replicate similar oil-well logging scenarios, with an aim to address the problem of replacement of conventional radioisotope sources commonly used in oil-well logging industries. An exploration tool similar to an oil-well logging tool was used to conduct experiments with water and sand as testing materials. The facility includes a 2500-gallon concrete test chamber with an aluminum pipe going horizontally through it. A machine-based 14.1 MeV deuterium-tritium neutron source as an alternative to conventional neutron sources was used. High energy neutrons assist in the investigation of a larger volume of material and also generate high energy inelastic scatter gamma rays, which provide useful information on composition. Experiments were performed with tap water and sand as a bulk testing material. Irradiation was done for one hour and results were obtained from a 3He neutron sensor, a BF3 neutron sensor, and two NaI gamma sensors placed at different locations within the exploration tool. Geant4, a Monte-Carlo based toolkit, was deployed on a high-performance computing system to simulate the entire experiment in order to benchmark the experimental responses obtained from the photon and neutron sensors. The facility was modeled in detail with accurate dimensions and material compositions. Materials such as tap water, high-density polyethylene, and aluminum metal were modeled with thermal neutron scattering cross-sections. The reference physics list QGSP_BIC_HP along with G4NDL and S(α,β) cross-sections were found to be appropriate for simulation of neutron interrogation experiments with neutron energies lower than 20 MeV. The experimental results obtained were successful in characterizing the bulk testing materials, and results obtained from Geant4 were found to be in good agreement with the experimental results in most cases.

Feasibility study of a compact neutron resonance transmission analysis instrument

Neutron Resonance Transmission Analysis (NRTA) uses resonant absorption of neutrons to infer the absolute isotopic composition of a target object, enabling applications in a broad range of fields such as archeology, enrichment analysis of nuclear fuel, and arms control treaty verification. In the past, NRTA involved large user facilities and complex detector systems. However, recent advances in the intensity of compact neutron sources have made compact neutron imaging designs increasingly feasible. This work describes the Monte Carlo (MC) based design of a compact epithermal NRTA radiographic instrument, which uses a moderated, compact deuterium-tritium neutron source and an epithermal neutron detector. Such an instrument would have a wide range of applications and would be especially impactful for scenarios such as nuclear inspection and arms control verification exercises, where system complexity and mobility may be of critical importance. The MC simulations presented in this work demonstrate accurate time-of-flight reconstructions for transmitted neutron energies, capable of differentiating isotopic compositions of nuclear material with high levels of accuracy. A new generation of miniaturized and increasingly more intense neutron sources will allow this technique to achieve measurements with greater precision and speed, with significant impact on a variety of engineering and societal problems.

Angular adaptive discrete ordinates method for deuterium–tritium neutron streaming experiments simulating narrow gaps

Shielding analyses for fusion reactors pose significant computational challenges, including strong anisotropic scattering and a combination of deep-penetration shielding and complex channeling. The discrete ordinates method is one of the most widely used techniques to solve the linear Boltzmann equation. We propose a regional angular adaptive algorithm to reduce angular discretization errors for shielding calculations. Angular adaptive refinement techniques are based on estimated local errors. The local errors are weighted by the importance distribution toward the computational goal, providing an appropriate goal-oriented adaptivity. The accuracy of this algorithm with FENDL/MG-3.0 libraries was investigated through analyses of streaming experiments with deuterium–tritium neutron sources at JAERI/FNS. From comparing reaction rates of measurements and calculations, the angular adaptive algorithm suffices in predicting the fast neutron field inside the gap. The solutions show that the algorithm can improve the accuracy of calculation results.

Applications of Serpent 2 Monte Carlo Code to ITER Neutronics Analysis

Nuclear analysis supporting the design and licensing of ITER is traditionally performed using MCNP and the reference model C-Model; however, the complexity of C-Model has resulted in the geometry creation and integration process becoming increasingly time-consuming. Serpent 2 is still a beta code; however, recent enhancements mean that it could, in principle, be applied to ITER neutronics analysis. Investigations have been undertaken into the effectiveness of Serpent for ITER neutronics analysis and whether this might offer an efficient modeling environment.An automated MCNP-to-Serpent model conversion tool was developed and successfully used to create a Serpent 2 variant of C-Model. A version of the deuterium-tritium plasma neutron source was also created. Standard reference tallies in C-Model for the blanket and vacuum vessel heating were implemented, and comparisons were made between the two transport codes assessing nuclear responses and computer requirements in the ITER model. Excellent agreement was found between the two codes when comparing neutron and photon flux and heating in the ITER blanket modules and vacuum vessel.Comparing tally figures of merit, computer requirements for Serpent were typically three to five times that of MCNP, and memory requirements were broadly similar. While Serpent was slower than MCNP when applied to fusion neutronics, future developments may improve this, and Serpent offers clear benefits that will reduce analyst time, including support for meshed geometry, robust universe implementation that avoids geometry errors at the boundaries, and mixed geometry types. Additional work is proceeding to compare Serpent against experiment benchmarks relevant for fusion shielding problems. While further developments are needed to improve variance reduction techniques and reduce simulation times, this paper demonstrates the suitability of Serpent to some aspects of ITER analysis.

A Study of the Fast Neutron Response of a Single-Crystal Diamond Detector at High Temperatures

The fast neutron response of a single-crystal diamond (SCD) detector was studied up to 300 °C. The SCD detector was fabricated using a $5\,\,\text {mm}\times 5\,\,\text {mm}\times 0.4\,\,\text {mm}$ substrate with premetallization treatments for obtaining oxygen terminated surfaces. The electrical contacts to the detector were provided by Cr/Au metallization. The performance of the fabricated detector was investigated using leakage current versus bias voltage (I–V) measurements and using pulse height spectra measurements with 5.5 MeV alphas and 14 MeV fast neutrons. The leakage current was observed to be about 400–600 pA at a field of 1 V/ $\mu \text{m}$ . From the full-width half-maximum of the alpha peak at 5.5 MeV, an energy resolution of about 2.1 ± 0.1% was estimated. The fast neutron response measurements were carried out using a 14 MeV deuterium–tritium neutron source. No measurable degradation of the detector performance was observed up to the maximum studied temperature of 300 °C. The neutron spectrum obtained at 300 °C showed a well-defined peak at 8.5 MeV due to the 12C(n, $\alpha)^{9}$ Be reaction which was the same as that observed at room temperature. Considering the total integrated counts, the detector sensitivity was observed to be $10^{-2}$ count per neutron. The results presented in this paper demonstrate for the first time the high temperature operation of an SCD detector up to 300 °C. These results also suggest that such SCD detectors could be used at higher temperatures for neutron diagnostics in the future fusion plasma facilities such as International Thermonuclear Experimental Reactor.

Performance testing of the neutron flux monitors from 10 keV to 1 MeV developed for BNCT: A preliminary study

The neutron flux monitors from 10keV to 1MeV designed for boron neutron capture therapy (BNCT) were experimentally tested with prototype monitors in an appropriate neutron field produced at the intense deuterium-tritium neutron source facility OKTAVIAN of Osaka University, Japan. The experimental test results and related analysis indicated that the performance of the monitors was good and the neutron fluxes from 10keV to 1MeV of practical BNCT neutron sources can be measured within 10% by the monitors.

Calculations to support JET neutron yield calibration: Modelling of neutron emission from a compact DT neutron generator

At the Joint European Torus (JET) the ex-vessel fission chambers and in-vessel activation detectors are used as the neutron production rate and neutron yield monitors respectively. In order to ensure that these detectors produce accurate measurements they need to be experimentally calibrated. A new calibration of neutron detectors to 14MeV neutrons, resulting from deuterium–tritium (DT) plasmas, is planned at JET using a compact accelerator based neutron generator (NG) in which a D/T beam impinges on a solid target containing T/D, producing neutrons by DT fusion reactions. This paper presents the analysis that was performed to model the neutron source characteristics in terms of energy spectrum, angle–energy distribution and the effect of the neutron generator geometry. Different codes capable of simulating the accelerator based DT neutron sources are compared and sensitivities to uncertainties in the generator's internal structure analysed. The analysis was performed to support preparation to the experimental measurements performed to characterize the NG as a calibration source. Further extensive neutronics analyses, performed with this model of the NG, will be needed to support the neutron calibration experiments and take into account various differences between the calibration experiment and experiments using the plasma as a source of neutrons.

Experimental study on the performance of an epithermal neutron flux monitor for BNCT

The performance of an epithermal neutron (0.5eV<En<10keV) flux monitor designed for boron neutron capture therapy (BNCT) was experimentally studied by using a prototype monitor in an appropriate neutron field at the intense deuterium-tritium neutron source facility OKTAVIAN of Osaka University, Japan. It was convinced from the experimental results that the developed monitor worked well and the epithermal neutron fluxes in BNCT neutron sources can be measured within 5% by the monitor.

3D imaging using combined neutron-photon fan-beam tomography: A Monte Carlo study

The application of combined neutron-photon tomography for 3D imaging is examined using MCNP5 simulations for objects of simple shapes and different materials. Two-dimensional transmission projections were simulated for fan-beam scans using 2.5MeV deuterium-deuterium and 14MeV deuterium-tritium neutron sources, and high-energy X-ray sources, such as 1MeV, 6MeV and 9MeV. Photons enable assessment of electron density and related mass density, neutrons aid in estimating the product of density and material-specific microscopic cross section- the ratio between the two provides the composition, while CT allows shape evaluation. Using a developed imaging technique, objects and their material compositions have been visualized.

Computational Study of Integrated Neutron/Photon Imaging for Illicit Material Detection

The feasibility of integration of photon and neutron radiography for nondestructive detection of illicit materials was examined. The MCNP5 code was used to model a radiography system consisting of accelerator-based neutron and photon sources and the imaging detector array, with an object under scrutiny placed between them. For this examination, the objects consisted of a matrix of low-Z and high-Z materials of various shapes and density. Transmission-radiography computations were carried out using 2.5-MeV deuterium-deuterium and 14-MeV deuterium-tritium neutron sources, and a 0.3-MeV photon source. The radiography tallies for both neutron and photon sources were modeled for the same geometry of the system. The photon-to- neutron transmission ratios were determined for each pixel of the detector array and utilized to identify the presence of specific materials in the radiographic images. By focusing on the inherent difference between neutron and photon interactions, it was possible to determine the shape and material composition of complex objects present within a pallet or a shipping container. The use of a single imaging array of scintillation detectors for simultaneous measurements of fast neutrons and photons is discussed, and its function in the dual neutron/photon radiography applications is addressed.

Compact deuterium-tritium neutron generator using a novel field ionization source

Active interrogation using neutrons is an effective method for detecting shielded nuclear material. A lightweight, lunch-box-sized, battery-operated neutron source would enable new concepts of operation in the field. We have developed at-scale components for a highly portable, completely self-contained, pulsed Deuterium-Tritium (DT) neutron source producing 14 MeV neutrons with average yields of 107 n/s. A gated, field ionization ion source using etched electrodes has been developed that produces pulsed ion currents up to 500 nA. A compact Cockcroft-Walton high voltage source is used to accelerate deuterons into a metal hydride target for neutron production. The results of full scale DT tests using the field ionization source are presented.

Measurements of (n, 2n) Reaction Cross Sections at 14 MeV for Several Nuclei

Measurements of (n, 2n) reaction cross sections to produce 84Rb, 86Rb, 88Y, 139Ce, 141Ce, 168Tm, 174m+gLu, 180mTa, 184m+gRe, 196n+gAu, and 237U were carried out in the range of 13.4 to 14.8 MeV. The samples were irradiated at various positions on the surface of a two-ring orientation instrument with a 20-cm radius centered at the deuterium-tritium neutron source. The 27Al(n,α)24Na reaction was used to monitor the neutron fluence in the target samples. The neutron energies of different directions were determined using the ratio of 89Zr to 92mNb specific activities induced in Zr and Nb foils by (n, 2n) reactions. The (n, 2n) products were measured using a calibrated Ge detector. Experimental uncertainty was within ±4%. The results are presented and compared with the results of references and ENDF/B-VII.0 evaluations. Cross sections of 85Rb(n, 2n)84Rb, 140Ce(n, 2n)139Ce, and 89Y(n, 2n)88Y were in good agreement with those of ENDF/B-VII.0 evaluations; however, disagreements were observed for 175Lu(n, 2n)174Lu and 142Ce(n, 2n)141Ce between the present data and the ENDF/B-VII.0 evaluation file. The present data showed improvement in accuracy in comparison with previously reported data; therefore, the results should be offered for evaluations of neutron (n, 2n) cross sections.

Effects of Fusion Produced Neutrons on Semiconductor Detectors

Detailed plasma-physics investigations by the use of x-ray-tomography data supported by the fundamental theoretical studies of x-ray-detector responses enhance the importance of x-ray diagnostics for fusion-plasma analyses. However, degradation in responses of semiconductor x-ray detectors after fusion-produced neutron exposure still remains one of the most serious problems in recent fusion experiments even at this time. For the purpose of investigating and characterizing neutron effects on semiconductor x-ray detectors, detection characteristics of n-type silicon semiconductor detectors which are similar to those utilized for x-ray-tomography detectors in the Joint European Torus (JET) tokamak, are studied by the use of synchrotron radiation from a 2.5-GeV positron storage ring at the Photon Factory. The fusion neutronics source (FNS) of Japan Atomic Energy Research Institute is employed as well-calibrated deuterium-tritium (D-T) neutron source with fluences from 1013 to 1015 neutrons/cm2 onto these semiconductor detectors. Degradation in x-ray responses with increasing neutron fluences has been reported; however, our recent detailed investigations of detector responses show nonlinear dependence as a function of the neutron fluence.

Optimization of Beam-Shaping Assemblies for BNCS Using the High-Energy Neutron Sources D-D and D-T

Boron neutron capture synovectomy is a novel approach for the treatment of rheumatoid arthritis. The goal of the treatment is the ablation of diseased synovial membranes in articulating joints. The treatment of knee joints is the focus of this work. A method was developed, as discussed previously, to predict the dose distribution in a knee joint from any neutron and photon beam spectra incident on the knee. This method is validated and used to design moderators for the deuterium-deuterium (D-D) and deuterium-tritium (D-T) neutron sources. Treatment times >2 h were obtained with the D-D reaction. They could potentially be reduced if the 10B concentration in the synovium was increased. For D-T neutrons, high therapeutic ratios and treatment times <5 min were obtained for neutron yields of 1014 s-1. This treatment time makes the D-T reaction attractive for boron neutron capture synovectomy.

Development of 14 MeV neutron spectrometer for fusion experimental reactor

A 14 MeV neutron spectrometer suitable for an ITER-like fusion reactor is conceptually designed on the basis of a recoil proton counter telescope method in oblique scattering geometry. To verify the principle and basic performance of the detector concept, preliminary experiments are made for a prototype detector system, where the detector responses are measured to 14 MeV neutrons produced from an accelerator deuterium-tritium (DT) neutron source and compared with the calculated ones. The comparison results show reasonably good agreement and demonstrate the possibility of energy resolution of 2.5% in full width at half maximum for 14 MeV neutron spectrometry.

Experimental Demonstration of Differing Impacts of Pulsed and Continuous Operation of a Deuterium-Tritium Neutron Source on Induced Radioactivity in the Context of ITER

The work reported herein was conducted in response to an ITER Task to demonstrate experimentally that pulsed and continuous operations of a D-T neutron source lead, in general, to differing impacts...

Neutronics Integral Experiments of Annular Blanket System Simulating Tokamak Reactor Configuration

Neutronics experiments on annular blanker systems that use a pseudoline source are performed. Deuterium-tritium neutron sources are set at the center of the inner cavity of the annular blanket system, and the pseudoline source is obtained by oscillating the annular blanket system back and forth in a 2-m span. Three annular blanket configurations are examined: the reference blanket, a blanket covered with 25-mm-thick graphite armor, and an armor blanket with a large opening (376 x 425.5 mm). The neutronics parameters of tritium production rate, neutron spectrum, and activation reaction rate are measured with specially developed techniques, including a multidetector data acquisition system, a spectrum weighting function method, and a ramp-controlled high-voltage system. Measured parameters are compared among three different configurations of the experimental system and also with the results of a closed geometry with a point source. A calculation with the GMVP Monte Carlo code that uses the JENDL-3 nuclear data library is performed and shows agreement within 10%. The current experiment provides unique data for a higher step of benchmark to test the ability of neutronics design calculations for a realistic tokamak reactor. 31 refs., 20 figs., 7 tabs.