Fusion Neutron Research Articles

Deterministic method for calculating fusion neutron emission probability distribution in tokamak through mesh mapping

The high-fidelity 3D neutronics analysis of fusion systems has been enabled by advanced computer-aided design (CAD)-based simulation tools, and is being driven by an increasing desire to couple the analysis results with other physical field analyses. With increase in the level of output detail, the effects of the fusion neutron source spatial distribution have become increasingly important. The current fine-resolution neutron source model was established based on the Monte Carlo (MC) method, which is a time-consuming approach to reduce the deviations at low-intensity positions. This work presents a deterministic method to calculate the fusion neutron emission probability (FNEP) distribution in a tokamak machine with high resolution. In this deterministic method, the mesh-mapping technique is utilized to transfer neutron intensity data from an irregular (a, φ) mesh to a structured or unstructured (R, Z) mesh. The intensity data on (R, Z) mesh can be used directly by a deterministic neutronics code, or further be integrated and normalized in every grid to obtain the FNEP for an MC neutronics code. This deterministic scheme was implemented in the code EDITON, and the neutron source map can be generated subsequently by this code. EDITON exhibited high efficiency, and its accuracy was verified by comparing the obtained FNEP distribution and neutron wall loading (NWL) with the corresponding data from the TRANSGEN code, which is a tokamak fusion neutron spatial distribution calculation code based on the MC method.

1-Hz Bead-Pellet Injection System for Fusion Reaction Engaged by a Laser HAMA Using Ultra-Intense Counter Beams

The injection and engagement of pellets using laser beam irradiation is one of the key technologies to realize a laser-driven inertial fusion energy (IFE) reactor. We irradiated ultra-intense laser (11 TW: 0.6 J/110 fs 2 beams with a focal intensity of 510 W/cm) in counter configuration on flying 1-mm-diameter deuterated polystyrene beads beyond 600 pellets on an average at 1 Hz and 10 min per cycle for 4 years. An injection system delivers pellets with free-fall that consists of a header for pellet delivery by disk rotation and a detection unit for synchronizing the motion of a pellet for laser engagement in time. During laser irradiation, the free-falling pellet placement was at Δx = 1 mm, Δy = 0.4 mm on a plane perpendicular to the falling direction, and Δz = 0.1 mm in the falling direction at the moment of laser irradiation. Using a two-directional probe shadowgraph system, we succeeded in aligning the pellet-falling position with a laser engagement probability greater than 70%; the probability improved from the previous experiments wherein the probabilities were less than 20%. As a result, the shot probability is 27% for gamma-ray generation resulting from ultra-intense laser-matter interactions and 22% for detection of signals corresponding to fusion neutrons with a maximum yield of 4 10 n/shot. The neutron reaction induced from an integrated system of pellet injector and laser is a decisive step in the research and development of an IFE reactor.

Velocity correction for neutron activation diagnostics at the NIF.

The velocity distribution of the hotspot in an inertial confinement fusion implosion changes the energy spectra of fusion neutrons emitted from the experiment as a function of viewing angle. These velocity-induced spectral changes affect the response of neutron activation diagnostics (NADs) positioned around the experiment and must be accounted for to correctly extract information about areal density (ρR) asymmetry from the data. Three mechanisms through which average hotspot velocity affects NAD activation are addressed: change in activation cross section due to the Doppler shift of the mean neutron energy, kinematic focusing of neutron fluence, and change in the scattering cross section due to the Doppler shift. Using the hotspot velocity inferred from neutron time-of-flight measurements of D-T and D-D fusion neutrons, the hotspot velocity is shown to account for the observed NAD activation asymmetry in a calibration shot with negligible fuel ρR. A robust method to evaluate uncertainties in spherical-harmonic fits to the NAD data due to the velocity correction and detector uncertainty is discussed.

A fused silica Cherenkov radiator for high precision time-of-flight measurement of DT and neutron spectra (invited).

A fused silica Cherenkov radiator has been implemented at the National Ignition Facility to provide a new high precision measurement of the time-of-flight spectrum of 14.1 MeV DT fusion neutrons. This detector enables a high precision (<30 ps) co-registered measurement of both a thresholded γ-ray and a neutron spectrum on a single record. Other methods typically require γ and neutron signals to be co-registered via other diagnostics and/or dedicated timing experiments. Analysis of the co-registered γ and neutron signals allows precise extraction of the mean neutron energy and bulk hot-spot velocity, both of which were not possible with prior scintillator technologies. Initial measurements demonstrate the feasibility of this measurement and indicate that combined detection of neutrons and γ-rays on multiple lines-of-sight should enable the bulk vector velocity of the implosion hot-spot to be determined to ≈5 km/s and reduced uncertainty in the spectral width ≈0.1 keV.

Application of the Denovo Discrete Ordinates Radiation Transport Code to Large-Scale Fusion Neutronics

Fusion energy systems pose unique challenges to the modeling and simulation community. These challenges must be met to ensure the success of the ITER experimental fusion reactor. ITER’s complex systems require detailed modeling that goes beyond the scale of comparable simulations to date. In this work, the Denovo radiation transport code was used to calculate neutron fluence and kerma for the JET streaming benchmark. This work was performed on the Titan supercomputer at the Oak Ridge Leadership Computing Facility. Denovo is a novel three-dimensional discrete ordinates transport code designed to be highly scalable. Sensitivity studies have been completed to examine the impact of several deterministic parameters. Results were compared against experiment as well as the MCNP and Shift Monte Carlo codes.

A study on neutron emission from a cylindrical inertial electrostatic confinement device

The adaption of new generation portable neutron sources has been increasingly marked in a wide range of research fields compared to the large-scale neutron generators. In this context, we have successfully demonstrated some essential parameters required for the emission of 2.45 MeV DD fusion neutrons from a steady state portable linear neutron source based on inertial electrostatic confinement scheme. The parameters that control the production of neutrons are the working pressure of the fuel gas, applied voltage, measured current and cathode geometries. The neutrons emitted from the source are confirmed using neutron monitor, bubble dosimeters, nuclear track detectors, and He-3 proportional counter. Presently, the device produces neutrons up to the order of ∼ 106 n/sec at discharge voltage ranging from -60 kV to -80 kV, and discharge current of 20 mA to 30 mA.

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.

Algorithmic Improvements to MCNP5 for High-Resolution Fusion Neutronics Analyses

Neutronics analyses of the ITER experimental fusion reactor rely on increasingly complex geometry models and estimates of energy-dependent neutron flux and radiation dose-rate distributions generated at ever higher resolutions. There are significant practical challenges with applying the Monte Carlo N-Particle (MCNP) continuous-energy transport code to high-resolution analyses. For models consisting of more than 100 000 surfaces and cells, geometry initialization can take several hours, thus slowing down model integration and transport analysis efforts. In multithreaded simulations, the amount of memory consumed by superimposed mesh tally data increases in proportion to the number of threads. This behavior limits either the tally resolution or the number of processor cores that can be utilized in the simulation. This paper describes algorithmic improvements that were implemented in a modified version of MCNP5 to overcome these limitations. These improvements are referred to as the Oak Ridge National Laboratory Transformative Neutronics (ORNL-TN) upgrade. A comparison of the performance and memory usage of both MCNP5 and ORNL-TN on several relevant fusion neutronics models is presented. In these tests and in actual high-resolution neutronics analyses, ORNL-TN reduces geometry processing times from hours to a few seconds and increases in-memory mesh tally capacity from the order of 108 to 1010 space-energy bins.

Uranium fusion implosion capsules see improved performance over gold

Bare depleted uranium hohlraum capsules exhibit double the fusion neutron yield and considerably reduced X-ray flux in inertial confinement fusion experiments compared to commonly used gold.

Fusion Neutrons: Tritium Breeding and Impact on Wall Materials and Components of Diagnostic Systems

A concise overview is given on the impact of fusion neutrons on various classes of materials applied in reactor technology: plasma-facing, structural and functional tested for tritium production and for diagnostic systems. Tritium breeding in the reactor blanket, fuel cycle and separation of hydrogen isotopes are described together with issues related to primary (tritium) and induced radioactivity. Neutron-induced damage and degradation of material properties are addressed. Material testing under neutron fluxes and safety issues associated with handling components in the radioactive environment are described. A comprehensive list of references to monographs and research papers is included to help navigation in literature.

First demonstration of improved capsule implosions by reducing radiation preheat in uranium vs gold hohlraums

In indirectly-driven Inertial Confinement Fusion (ICF) implosions, supra-thermal M-band (&amp;gt;2 keV) radiation from principally 4–3 resonance line transitions generated during laser irradiation at the peak power of Au hohlraum walls can preheat the fusion capsule and reduce compressional pressure. Higher Z, un-lined depleted uranium (DU) hohlraums were used for the first time in ICF implosions on the National Ignition Facility (NIF) to reduce M-band radiation levels while keeping the total radiation flux similar to Au hohlraums. First implosions in DU demonstrate an increase in in-flight density (+15%) of high density carbon capsules, and hence in stagnated hot spot temperature (+15%), hot spot x-ray (+200%) and fusion neutron yields (+100%) compared to Au hohlraums. We show analytically that these changes are consistent with the observed 40% reduction in M-band x-ray flux in DU, and are in agreement with 2D hydrodynamic simulations. This result had a major impact on ICF research on the NIF where a significant fraction of high neutron yield implosions are currently using un-lined DU hohlraums.

Tests of refractory and high-porous metals in plasma device at National Research University MPEI

The plasma device at the National Research University “MPEI” has been constructed within the scope of the National controlled fusion program to test materials for fusion reactors, namely, fusion neutron sources FNS, DEMO and the international fusion reactor ITER. The device is a plasma trap with a linear multicusp configuration of the magnetic field and stationary plasma discharge that provides high-power plasma thermal load on the samples under testing which are made of materials of the fusion reactor first wall. It is planned to develop a new technology for creating highly porous structure on the surfaces of refractory metals, including the so-called tungsten “fuzz” with diameters of pores and nanofibers of ∼50 nm, which is of considerable interest for nuclear, power engineering and biomedical technologies.

Thin foil proton recoil spectrometer performance study for application in DT plasma measurements.

The Thin foil Proton Recoil (TPR) technique has previously been used for deuterium-tritium fusion neutron diagnostics [N. P. Hawkes et al., Rev. Sci. Instrum. 70, 1134 (1999)] and is one of the candidates put forward for use in ITER as part of the high resolution neutron spectrometer (HRNS) system [E. A. Sundén et al., Nucl. Instrum. Methods Phys. Res., Sect. A 701, 62 (2013)]. For ITER, the neutron spectrometer's main purposes are to determine the fuel ion density ratio as well as the ion temperature in DT plasma. This work focuses on testing the capability of a proton telescope detector intended for use as part of the TPR spectrometer. The proton telescope has been tested using proton energies in the range of 3-8 MeV. The experimental results cover energy calibration, resolution estimation, and testing the spectrometer's capability to perform background separation using ΔE - E energy cuts. In addition, spectrometer performance in terms of signal to background ratios for ITER-like DT plasma conditions is estimated using Monte-Carlo simulations. Results show that the TPR spectrometer geometry dominates in determining the energy resolution and the ΔE - E energy cuts will significantly reduce the background. In addition, the estimated spectrometer count rates in ITER-like conditions fall below 20 kHz per detector segment.

Review of Neutron Diagnostics Based on Fission Reactions Induced by Fusion Neutrons

Neutrons are an exceptional diagnostic signature in magnetic fusion experiments with high temperature plasma. A part of neutron detection methods relies on detecting nuclear reactions products, initiated by neutrons. Among different possibilities, the neutron-induced fission reactions can be used. These methods use the specific properties of the radioactive decay phenomena induced by neutrons of heavy elements like uranium or thorium. The neutron-induced fission reactions produce energetic fission fragments that can be detected in a special type of ionization chambers that has one of electrode covered with fissionable material. A fission reaction may be also adapted to neutron detection by activation method. A dedicated target-sample made from fissionable isotope (like 235U, 238U, 232Th), irradiated by neutrons, is the source of prompt and delayed neutrons. The decay curve of delayed neutrons can be registered almost immediately after irradiation by fusion neutrons, with a delay resulting exclusively from the time of transport of the samples to the measuring set-up. The yield of a fusion neutron source can be calculated from that measurement. Fission phenomena and fissionable materials are applicable in plasma neutron diagnostics. The lecture gives the introduction to the nuclear fission process and the particular attention is done to the phenomenon and physics of delayed neutrons which are generated during the fission reaction. Examples of delayed neutrons activation set-up and fission chambers dedicated for tokamaks are presented.

HiperFer, a reduced activation ferritic steel tested for nuclear fusion applications

Materials are the most urgent issue in nuclear fusion research. Besides tungsten, steels are considered for unifying functional and structural materials due to their cost and mechanical advantages over tungsten. However, the fusion neutrons impose a strong constraint on the ingredients of the steel in order to avoid long lasting activation, while the material has to pertain sputtering resistance, low hydrogen retention, and long-term mechanical stability. In this proof-of-principle, we demonstrate the interesting properties of the new material HiperFer (High performance Ferrite) as a material suitable for fusion applications.The investigation covers neutron activation modelled by FISPACT-II, plasma sputtering and deuterium retention experiments in PSI-2, thermo-mechanical properties and component modelling. The material was found to feature a low nuclear inventory. Its sputtering yield reduces due to preferential sputtering by a factor 4 over the PSI-2 D2 plasma exposure with possible reductions of up to 70 indicated by SD.Trim.SP5 modelling. The exposure temperature shows a strong influence on this reduction due to metal diffusion, affecting layers of 1 µm in PSI-2 at 1150 K exposure for 4 h. Deuterium retention in the ppm range was found under all conditions, together with ∼10 ppm C and N solubility of the ferritic material. The creep and cyclic fatigue resistance exceed the values of Eu-97 steel. As an all HiperFer component, heat loads in the order of 1.5 MW/m² could be tolerated using water-cooled monoblocks. In conclusion, the material solves several contradictions present with alternative reduced-activation steels, but its applications temperatures >820 K also introduce new engineering challenges.

Multifunctional Neutronics Calculation Methodology and Program for Nuclear Design and Radiation Safety Evaluation

Advanced nuclear systems, such as fusion systems, generally have features of large size, complex structures, spatially heterogeneous distribution of components and materials, and high energy and high flux, as well as a wide and complex energy spectrum of neutrons. Compared with traditional nuclear systems, these features have brought unprecedented challenges to neutronics design and analysis. To confront these challenges, the FDS Team has made significant progress in the development of neutronics methods and the comprehensive simulation code Super Multi-functional Calculation Program for Nuclear Design and Safety Evaluation (SuperMC). Furthermore, the FDS Team has been developing the High Intensity D-T Fusion Neutron Generator (HINEG) and has performed a series of neutronics experiments. Based on the developed methods, codes, and facility, a series of fusion designs and analyses has been carried out, including the design of FDS series reactors as well as the ITER neutronics analysis.

Cyclooxygenase‑2‑mediated upregulation of heme oxygenase 1 mitigates the toxicity of deuterium‑tritium fusion radiation.

Utilizing the energy released from the nuclear fusion of deuterium with tritium (D-T) may be an important method of supplying energy in the future. The ionizing radiation emitted from nuclear fusion is a potential health risk to humans, including scientists who are currently performing nuclear fusion experiments and the employees of fusion nuclear plants, in the future. However, there have been few reports on the biological effects of fusion radiation. In the present study, using the High Intensity D-T Fusion Neutron Generator, the DNA damage and its regulation in normal human fibroblasts exposed to fusion radiation were investigated. Heme oxygenase 1 (HO-1), which is reported to induce anti-inflammatory activity, was upregulated in the irradiated cells. Pretreatment with the HO-1 inhibitor, protoporphyrin IX zinc (II), exacerbated double strand break formation following exposure to fusion radiation. The expression of cyclooxygenase-2 (COX-2) contributed to the upregulation of HO-1, as demonstrated by the result that its inhibitor, NS-398, inhibited the induction of HO-1 in irradiated cells. It was further clarified that the ataxia telangiectasia mutated DNA damage response was activated and it stimulated the phosphorylation of p38 mitogen-activated protein kinase, which was responsible for the upregulation of COX-2 and HO-1. These results provide novel information on fusion radiation-induced biological effects and potential targets for decreasing the associated health risks.

A study regarding the optimal radial build of a low-aspect-ratio tokamak fusion system

System parameters and the optimal radial build of a low-aspect-ratio (LAR) tokamak fusion system were found through the coupled analysis of a tokamak system and neutron transport. In a configuration with only an outboard breeding blanket, the minimum major radius to produce a given fusion power was determined by the shielding requirements and the magnetic field at the toroidal field (TF) coil. With a confinement enhancement factor H = 1.4, Q > 10 was possible for fusion power greater than 800 MW with an aspect ratio of A = 1.5; however, Q > 10 was possible for fusion power greater than 1800 MW with an aspect ratio of A = 2.0. The outboard radial build was determined by the tritium breeding and shielding requirements. The tritium breeding capability of blanket concepts proposed for testing in the International Thermonuclear Experimental Reactor (ITER) was evaluated by varying the outboard blanket thickness and the degree of lithium-6 (Li-6) enrichment. Cases with a smaller aspect ratio exhibited better performance since the number of fusion neutrons that contributed to tritium breeding were larger than the case with a larger aspect ratio. Among the blanket concepts, a helium (He)-cooled solid breeder (HCSB) concept showed the best tritium breeding capability and thus allowed for a smaller system size.

Silicon carbide PIN diode detectors used in harsh neutron irradiation

Silicon carbide (SiC) neutron detectors based on PIN diode with Ti/Au electrode structure have been successfully developed and considered to be a good option for neutron detection related to harsh radiation environment, but these detectors suffered inevitable irradiation damage and performance degradation, ultimately damaged the applications severely. Unfortunately, the systematic research focusing on the degradation and irradiation defects is lack, and the mechanism of degradation is not clear yet. In this paper, these issues are carefully studied: SiC detectors were irradiated by neutrons emitted from fusion neutron generator and nuclear reactor with fluence ranging from 1 × 1014 n/cm2 to 2 × 1016 n/cm2, then the performance degradation and neutron induced defects were carefully investigated. It is concluded that the SiC detectors could endure neutron irradiation as least 1016 n/cm2. Although neutron irradiation could induce breakdown of PN junction, decrease of charge collection efficiency and worsening of energy resolution, all the detectors are still alive; furthermore, they could still acquire integrated response spectra to alpha particles with neglectable increase of dark current at neutron fluence of 2 × 1016 n/cm2. The di-carbon antisite defects and vacancy-type defects induced by neutron irradiation should be responsible for the performance degradation, according to low temperature photoluminescence and positron annihilation Doppler broadening measurements. Possible degradation mechanism was discussed as well.

TLD calibration for neutron fluence measurements at JET fusion facility

Measurements of neutron streaming through penetrations in biological shields are being carried out at JET fusion device by means of thermoluminescence (TL) detectors with the objective to validate the neutronics codes and nuclear data widely applied in ITER nuclear analyses in a real fusion environment. TLDs response due to the neutron component of the radiation field is related to the neutron fluence in a well-defined neutron energy spectrum. Therefore, a TLDs calibration in real fusion radiation fields is necessary to allow neutron fluence from TL measurements at JET to be more precisely calculated. Hence, a MCP-N and MCP-7 TLDs produced at the IFJ PAN in Kraków were calibrated at the ENEA facilities of Frascati and Casaccia laboratories. The obtained results have been analysed and new calibration factors are proposed. The detection system based on TLDs developed and calibrated for JET experiments can then be generally applied not only to fusion neutron fields, but also to ITER to monitor the neutron fluence outside the biological shield.