Neutron Yield Research Articles

Generalized plasma focus problem and its application to space propulsion

Space propulsion is unique among many proposed applications of the dense plasma focus in being critically dependent on the availability of a scaling theory that is well-grounded in physics, in conformity with existing experimental knowledge and applicable to experimentally untested configurations. This paper derives such a first-principles-based scaling theory and illustrates its application to a novel space propulsion concept, where the plasma focus sheath is employed as a power density amplifying mechanism to transport electric energy from a capacitive storage to a current-driven fusion load. For this purpose, a Generalized Plasma Focus problem is introduced and formulated. It concerns a finite, axisymmetric plasma, driven through a neutral gas at supersonic speed over distances much larger than its typical gradient scale length by its azimuthal magnetic field while remaining connected with its pulse power source through suitable boundaries. The Gratton-Vargas equation is rederived from the scaling properties of the equations governing plasma dynamics and solved for algebraically defined initial (insulator) and boundary (anode) surfaces. Scaling relations for a new space propulsion concept are derived. This consists of a modified plasma focus with a tapered anode that transports current from a pulsed power source to a consumable portion of the anode in the form of a hypodermic needle tube continuously extruded along the axis of the device. When the tube is filled with deuterium, the device may serve as a small-scale version of magnetized liner inertial fusion (MAGLIF) that could avoid failure of neutron yield scaling in a conventional plasma focus.

ON THE PHYSICAL PARAMETERS OF THE STELLARATOR – SOURCE OF NEUTRONS WITH NEOCLASSICAL LOSSES, TAKING INTO ACCOUNT RECYCLING AND BOHM LOSSES

The results of calculations of the physical parameters of the stellarator used as a neutron source for initiating nuclear reactions in a blanket of a hybrid reactor are presented. The calculations were carried out assuming the implementation of neoclassical transport processes, taking into account recycling and Bohm losses. The crushing effect of Bohm diffusion on the results of neutron yields is shown.

Neutron and gamma-ray signatures for the control of alpha-emitting materials in uranium production: A Nedis2m-MCNP6 simulation

The radiations emitted via 19F(α,nγ), 19F(α,pγ), and 19F(α,α′γ) reactions in compounds containing alpha-emitting elements and fluorine are used for non-destructive neutron and gamma-ray analyses in monitoring the contents of nuclear fissile materials during the manufacture, processing, and storage of nuclear fuels. As far as the literature survey confirms, the data available on the yield of (α,n) reaction on fluorine, which depends on alpha particle energy, have significant dispersion. For the accompanying gamma-rays, on the other hand, no reliable absolute data exist for the production yield of gamma-rays per number of interacting alpha particles or emitted nucleons as a function of alpha particle energy. Therefore, to improve the monitoring and control of nuclear materials and to ensure the safety of production processes, the present study was undertaken, where the neutron yields and energy spectra were calculated using the Nedis-2m program taking into account the updated values of the total cross-section for (α,n) reactions on 19F. The resulting Nedis-2m input datasets were used in the MCNP6 code to calculate the leakage multiplication factor and neutron energy spectrum.

GEANT4 dose estimations of solar protons: Al and PMMA-Bi2O3 shielding for space exploration

Adverse effects of long-term exposure to galactic cosmic radiation (GCR) pose a clear obstacle to future space exploration programs. In addition to GCR we have solar particle radiation. We simulated the latter using a scaled fluence profile of solar protons taken from a literature study that comprises about three solar cycles. The model is a three-layer stack that includes shielding material and muscular tissue. Our simulation strategy uses protons as precursor radiation of neutrons. Subsequently, the shield is adjusted for thickness, dictated by an average depth at which neutrons are created through various processes during the simulation. Neutrons are then energy-binned and a corresponding neutron flux is simulated. Particles generated during the second phase of the simulation, i.e. by neutrons, are then counted toward absorbed dose within the muscular tissue layer. Clearly, the dynamics of the process is not captured by the simulation, nevertheless an overview of neutron yield can be estimated and the absorbed dose. The objective is to provide some insight about the effect of the new composite shield, PMMA-Bi2O3, that has an intrinsic capability for gamma dose reduction, compared to a more traditional aluminum shield.

Experimental study of spatial resolution of MCPs for compact high-resolution neutron radiography system

Neutron-sensitive microchannel plates (MCPs) have been widely used as detectors in neutron radiography systems due to their high neutron detection efficiency and superior spatial resolution. These MCPs are typically employed in large-scale facilities with high neutron flux and large L/D ratios, allowing them to achieve excellent spatial resolution. However, it was uncertain whether the MCPs’ high-resolution advantage could be fully utilized in compact radiography systems with limited neutron yield and lower L/D ratios. To investigate this, thermal neutron imaging experiments were conducted in this paper. The results showed that MCPs can offer a 0.25 mm spatial resolution, compared to the 0.3 mm of a 6LiF/ZnS detector. Additionally, a centroid algorithm imaging technique was utilized to further enhance the MCPs’ spatial resolution up to 0.2 mm. This research on MCPs can also be used as a reference for other neutron radiography systems with limited L/D ratios.

Reaching a burning plasma and ignition using smaller capsules/Hohlraums, higher radiation temperatures, and thicker ablator/ice on the national ignition facility

In indirect-drive implosions, the final core hot spot energy and pressure and, hence, neutron yield attainable in 1D increase with increasing laser peak power and, hence, radiation drive temperature at the fixed capsule and Hohlraum size. We present simple analytic scalings validated by 1D simulations that quantify the improvement in performance and use this to explain existing data and simulation trends. Extrapolating to the 500 TW National Ignition Facility peak power limit in a low gas-fill 5.4 mm diameter Hohlraum based on existing high adiabat implosion data at 400 TW, 1.3 MJ and 1 × 1016 yield, we find that a 2–3 × 1017 yield (0.5–0.7 MJ) is plausible using only 1.8 MJ of laser energy. Based on existing data varying deuterium–tritium (DT) fuel thickness and dopant areal density, further improvements should be possible by increasing DT fuel areal density, and hence confinement time and yield amplification.

Measurement of thick target neutron yield from 80.5 MeV/u 12C incidence on Be, C, W, and Pb targets

Thick target neutron yield produced from 80.5 MeV/u 12C ions stopping in the thick Be, C, W, and Pb targets were measured at 0°, 37°, 60° using liquid scintillation detector and the time of flight method. The experimental data were compared with those calculations using the Liege IntraNuclear-Cascade (INCL) implemented in Geant4 and the JAERI Quantum Molecular Dynamics 2.0 (JQMD-2.0) model followed by the Generalized Evaporation Model (GEM) implemented in PHITS, and default model implemented in FLUKA. The results of the three simulation codes with different physics models are similar for beryllium and carbon targets and in good agreement with the experimental data. For tungsten and lead targets, the calculation results from PHITS and FLUKA successfully reproduced the neutron spectra. While the results from GEANT4 overestimate the experimental data above 50 MeV at 0° and significantly underestimated experimental data in the 15–60 MeV energy range at 0°, 37° and 60°. In general, the simulation results of PHITS with the JQMD-2.0 physical model are in good agreement with the experimental data for all targets.

Estimation of the yield and angular distribution of neutrons for calculating the biological shielding of reconstructed heavy ion accelerators with energies from 1 to 6 MeV/u

A simple analytical formulas for estimation of the yield and angular distribution of neutrons from thick targets would be useful for calculating the biological protection of heavy ion accelerators with energy (1–6) MeV/nucleon. The results of the calculation with this formulas were compared with those obtained using the LISE++ and FLUKA programs as well as with the available experimental data. As a result, it can be stated that the proposed calculation method can be used for quick assessment. The deviation of the results of such calculations from more accurate ones does not exceed a factor of two, which is comparable with the deviations in the calculation of biological shielding due to a variety in the protection properties of the materials used or in their thickness.

Simulation and assessment of material mixing in an indirect-drive implosion with a hybrid fluid-PIC code

Hybrid fluid-PIC simulations aimed at a better understanding of the implosion physics and the material mixing into the hot spot are described. The application of a hybrid fluid-PIC code is motivated by the difficulty of modeling the material mixing by the commonly used radiation hydrodynamic simulations. Hybrid fluid-PIC techniques, which treat the ions with the traditional particle-in-cell method, and electrons with a massless fluid, are more adaptable to handle the heating of DT fuel through PdV work and the material mixing near the DT ice-gas interface and ablator-fuel interface of a compressed capsule. During implosion shock convergence, significant reactant temperature separation and a noticeable amount of material mixing are observed, both of which have important consequences for estimating neutron yield and the understanding of implosions. Physical explanations for these phenomena are discussed, with the non-equilibrium effect in the hotspot and hydrodynamic instabilities at the interface as the likely explanation, respectively. The hybrid fluid-PIC method would be helpful to test the phenomenological fluid model describing the material mixing in ICF implosion.

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.

First results from third harmonic ion cyclotron acceleration of deuterium beams in EAST ion heating studies experiments

Recent Experimental Advanced Superconductivity Tokamak experiments have been dedicated to the studies of increasing ion temperature using Neutral Beam Injection (NBI) in synergy with radio-frequency waves in the Ion Cyclotron Range of Frequencies (ICRFs) in third harmonic. They showed that this scenario effectively accelerates fast deuterium ions and produces a larger tail of fast neutrons with E > 3.0 MeV, as measured by Neutron Energy Spectrum. Comparing to the scenario with 2.2 MW NBI alone, adding an addition of 0.3 MW ICRF power leads to an enhancement of fusion yield by 67% (from to ) and an increase of core ion temperature by . Besides, the simulation results of TRANSP show that the neutron yield produced by 0.2 MW ICRF in third harmonic with NBI synergy is roughly consistent with the experimental results. Meanwhile, the 80 keV fast ion tail generated by NBI can be heated up to 400 keV and 600 keV by the 0.2 MW and 1.0 MW ICRF with the synergistic effect, respectively. In addition, it is observed for the first time that this ICRF-NBI synergy leads to the excitation of fishbone instabilities with characteristic frequencies of (10–20) kHz. Furthermore, parameter scans of ICRF/NBI power and plasma density were performed for third harmonic ICRF and NBI synergetic heating schemes. It is found that the neutron yield increases nonlinearly as the ICRF power or plasma density increases.

Machine learning for detection of 3D features using sparse x-ray tomographic reconstruction.

In many inertial confinement fusion (ICF) experiments, the neutron yield and other parameters cannot be completely accounted for with one and two dimensional models. This discrepancy suggests that there are three dimensional effects that may be significant. Sources of these effects include defects in the shells and defects in shell interfaces, the fill tube of the capsule, and the joint feature in double shell targets. Due to their ability to penetrate materials, x rays are used to capture the internal structure of objects. Methods such as computational tomography use x-ray radiographs from hundreds of projections, in order to reconstruct a three dimensional model of the object. In experimental environments, such as the National Ignition Facility and Omega-60, the availability of these views is scarce, and in many cases only consists of a single line of sight. Mathematical reconstruction of a 3D object from sparse views is an ill-posed inverse problem. These types of problems are typically solved by utilizing prior information. Neural networks have been used for the task of 3D reconstruction as they are capable of encoding and leveraging this prior information. We utilize half a dozen, different convolutional neural networks to produce different 3D representations of ICF implosions from the experimental data. Deep supervision is utilized to train a neural network to produce high-resolution reconstructions. These representations are used to track 3D features of the capsules, such as the ablator, inner shell, and the joint between shell hemispheres. Machine learning, supplemented by different priors, is a promising method for 3D reconstructions in ICF and x-ray radiography, in general.

Investigation of the Dependence of the Yield of Neutrons and Protons of the DD Reaction from Ti and CVD Diamond on the Target Rotation Angle

Investigation of the Dependence of the Yield of Neutrons and Protons of the DD Reaction from Ti and CVD Diamond on the Target Rotation Angle

Investigation of the Dependence of the Yield of Neutrons and Protons of the DD Reaction from Ti and CVD Diamond on the Target Rotation Angle

Investigation of the Dependence of the Yield of Neutrons and Protons of the DD Reaction from Ti and CVD Diamond on the Target Rotation Angle

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.

Design and experimental validation of differential pumped vacuum system for HINEG windowless gas target

High Intensity Neutron Generator (HINEG) is a facility aiming for neutronics design validation, materials and components irradiation test, nuclear waste transmutation and nuclear technology applications, etc. A high-pressure windowless gas target system is adopted for HINEG to achieve a higher neutron yield. Therefore, a three-stage differential vacuum system has been developed to couple the pressure transition of 6 orders of magnitude to the 300 kV/50 mA high-intensity beam transport in a series of large-diameter pipes. A suitable pumping system has been developed to increases the target pressure which is used to generate D-T neutron yield of 2.1 × 1013 n/s. A windowless pressure transition from high gas pressure (1 × 103 Pa) in the gas target chamber to high vacuum (3 × 10−3 Pa) in the accelerator is reached. In this paper, a new differential pumped vacuum system is designed for HIENG. Innovatively, the newly designed differential vacuum system can not only meet the transport requirements of the high-current low-energy deuteron beam, but also couple the pressure transition of 6 orders of magnitude, which can solve the contradiction of the large-diameter pipe required for high-intensity beam transport and the small diameter pipe needed for the high-pressure difference.The experimental validation and design optimization of differential pumped vacuum system is introduced. The experiment results indicate the validity of the theoretical calculation. Compared to molecular pumps, roots pumps might be more suitable for windowless differential vacuum system coupled to large-diameter apertures and pipes. The design could be used for reference in other windowless gas target systems coupled to accelerator with high-current low-energy ion beam.

Laser-Driven Neutron Generation Realizing Single-Shot Resonance Spectroscopy

Experiments identify the mechanism that accelerates ions in a laser-driven neutron source (LDNS) as well as a scaling law for the neutron yield, key insights that move LDNS closer to practical neutron generation.

Benchmark shielding calculations for fusion and accelerator-driven sub-critical systems

We conducted benchmark calculations of neutron-shielding experiments for fusion applications such as (1) Neutron spectra in iron shields with 14 MeV neutrons; and (2) Leakage neutron spectra from various sphere piles by 14 MeV neutrons, and for Accelerator-Driven Sub-Critical system (ADS) such as (3) Neutron energy spectra after transmission through iron and concrete shields using a quasi-monochromatic neutron source, (4) Thick target neutron yield produced by high-energy heavy ion incidence on a thick target, and (5) Induced radio activity in concrete exposed to secondary particles produced by heavy-ion incident reaction on an iron target by using the Particle and Heavy Ion Transport code System (PHITS). For case (1), the calculated neutron energy spectra using the nuclear data libraries, JENDL-4.0, agreed with the experimental data in iron shields. For the cases (2) of Al, Ti, Cr, and As piles, the neutron energy spectra calculated with JENDL-4.0 agreed with the experimental results well. For the case (3), the neutron spectra calculated with JENDL-4.0/HE reproduced the experimental data because the proton data library of7Li can produce source neutrons for proton incident reactions. For the case (4), calculated results for thick target neutron spectra produced by a 400 MeV per nucleon carbon incident reaction on lead reproduced the experimental data well. For the case (5) with a neutron source produced by the 200–400 MeV per nucleon heavy-ion incident reaction, the calculated results of the reaction rate depth profiles of197Au (n, γ)198Au reactions reproduced the experimental results within a factor of 2. Overall, PHITS with nuclear data libraries reproduced the experimental data sufficiently well. Thus, PHITS has a potential application in the design of advanced reactor systems, such as fusion and ADS facilities. This experimental data are also useful in validating nuclear reaction models in other Monte Carlo codes and evaluating nuclear data libraries for advanced reactor systems.

1D/2D benchmark experiments of aluminum alloy irradiated by 14 MeV neutrons for ICF facilities

Aluminum alloy (5083) is one of the most important structural materials used in laser-driven Inertial Confinement Fusion (ICF) facilities, which will be exposed to high-yield 14 MeV neutrons and highly activated after the DT explosion. Reliable estimation of shutdown dose rate levels is of great importance for machine operation planning and to ensure the safety of people entering into target bay. Both 1D and 2D benchmark experiments were performed on a DT neutron accelerator at the Institute of Nuclear Physics and Chemistry (INPC) to validate the calculation codes and cross section libraries (JMCT with FENDL-3.1d and FISPACT with EAF-2007) used in the simulations of neutron transport and activation for aluminum alloy 5083. The total neutron yield was about 6 × 1014 in the experiments. Both radioactivity of foils placed inside the assembly and air kerma rate at given positions out of the assembly have been measured in 1D/2D experiments. Monte-Carlo code JMCT with FENDL-3.1d was used in both neutron and gamma-ray transport, and the inventory code FISPACT with EAF-2007 was used to calculate neutron activation. Results show that the relative error for air kerma rate in simulations and experiments is less than 20%, and radioisotopes give different contribution ratios on dose levels for 1D/2D experimental assemblies after DT neutron irradiation, especially for 24Na, 27Mg and 56Mn at the time within 16 h after neutron irradiated. Besides, simulations of radioactivity give good agreement with experimental results obtained for radioisotopes 24Na, 196Au, 57Ni and 58Co in both 1D and 2D experiments, and the relative error is less than ±16%.

Target cooling options for DARIA compact neutron source

The extensive heat release in the target is the primary limiting factor for a CANS neutron output. CANS DARIA has been chosen to operate using a 13 MeV proton beam providing up to 40 kW of power, which requires an effective target cooling solution. It was found that beryllium provides the best neutron yield while staying in solid state, which makes it the most effective option for the target material. With an optimal beryllium target thickness of 1.1 mm, the proton Bragg peak lies outside of the beryllium layer, but 9.21 MeV per incident proton are still dissipated inside the beryllium. Two cooling options are considered and analysed with PHITS calculations: multilayer targets and rotating targets. The use of proton beams with energies above 13 MeV on beryllium leads to tritium generation, which is not desirable. Any lower energy leads to a decreased neutron yield, but a simpler cooling solution. Therefore, an option to reduce the proton beam energy is also considered.