Fusion Neutron Flux Research Articles

Ion optical design of the magnetic proton recoil neutron spectrometer for the SPARC tokamak.

A magnetic proton recoil (MPR) neutron spectrometer is being designed for SPARC, a high magnetic field (BT = 12T), compact (R0 = 1.85m, a = 0.57m) tokamak currently under construction in Devens, MA, USA. MPR neutron spectrometers are versatile tools for making high fidelity abinitio calibrated measurements of fusion neutron flux spectra and have been used to infer fusion power, ion temperature, fuel ion ratio, and suprathermal fuel populations at several high performance fusion experiments. The performance of an MPR neutron spectrometer is in large part determined by the design of the magnetic field, which disperses and focuses recoil protons. This article details the ion optical design of a high-resolution MPR neutron spectrometer, including the amelioration of image aberrations due to nonlinear effects. An optimized design is presented that achieves ion optical energy resolution δE/E < 1% and focal plane properties that enable straightforward integration with the hodoscope detector array.

A Plasma-Window Enhanced Accelerator-Based Deuterium-Tritium Neutron Generator System

There is a known demand for a fusion prototypic neutron source capable of emulating the neutron-induced damage caused by fusion. If no such source is developed in a timely and economical manner, the use of fusion as a source of energy will be hindered by material selection and qualification. Presented here is one possible path toward the development of a fusion prototypic neutron source by enhancing an operational neutron generator platform with so-called plasma windows. The use of plasma windows addresses a weakness in the current design by improving the pressure differential between acceleration and the target regions. This improvement, combined with the use of multiple beamlines, represents the possibility of dramatically increasing the fusion neutron flux capabilities of such a system.

Maintaining the close-to-critical state of thorium fuel core of hybrid reactor operated under control by D-T fusion neutron flux

The results of full-scale numerical experiments of a hybrid thorium-containing fuel cell facility operating in a close-to-critical state due to a controlled source of fusion neutrons are discussed in this work. The facility under study was a complex consisting of two blocks. The first block was based on the concept of a high-temperature gas-cooled thorium reactor core. The second block was an axially symmetrical extended plasma generator of additional neutrons that was placed in the near-axial zone of the facility blanket. The calculated models of the blanket and the plasma generator of D-T neutrons created within the work allowed for research of the neutronic parameters of the facility in stationary and pulse-periodic operation modes. This research will make it possible to construct a safe facility and investigate the properties of thorium fuel, which can be continuously used in the epithermal spectrum of the considered hybrid fusion–fission reactor.

Calibration study for a “fission electron-collection” neutron detector

Abstract The “fission electron-collection” neutron detector (FECND) is a fission-based neutron detector with flat and fast responses, and is a promising choice for the diagnostics of fusion neutron flux and spectrum. Due to its low sensitivity, the detector output is weak and is disturbed easily in calibration experiments. The backgrounds in the calibration primarily come from scattered neutrons and the γ rays induced by neutron inelastic scattering. A symmetrical structure of FECND was proposed to suppress the output of γ rays, and its effectiveness was validated by Monte Carlo simulations. A method was presented to eliminate the impact of scattered neutrons by utilizing the solid angle effect and the energy difference between source neutrons and scattered neutrons. Through calibration experiments and analyses, a D-T sensitivity of (8.6 ± 0.2)×10−20 C·cm2 was obtained for the FECND.

Neutron diagnostics for the physics of a high-field, compact, Q ≥ 1 tokamak

Advancements in high temperature superconducting technology have opened a path toward high-field, compact fusion devices. This new parameter space introduces both opportunities and challenges for diagnosis of the plasma. This paper presents a physics review of a neutron diagnostic suite for a SPARC-like tokamak [Greenwald et al., 2018, https://doi.org/10.7910/DVN/OYYBNU]. A notional neutronics model was constructed using plasma parameters from a conceptual device, called the MQ1 (Mission Q ≥ 1) tokamak. The suite includes time-resolved micro-fission chamber (MFC) neutron flux monitors, energy-resolved radial and tangential magnetic proton recoil (MPR) neutron spectrometers, and a neutron camera system (radial and off-vertical) for spatially-resolved measurements of neutron emissivity. Geometries of the tokamak, neutron source, and diagnostics were modeled in the Monte Carlo N-Particle transport code MCNP6 to simulate expected signal and background levels of particle fluxes and energy spectra. From these, measurements of fusion power, neutron flux and fluence are feasible by the MFCs, and the number of independent measurements required for 95% confidence of a fusion gain Q ≥ 1 is assessed. The MPR spectrometer is found to consistently overpredict the ion temperature and also have a 1000× improved detection of alpha knock-on neutrons compared to previous experiments. The deuterium-tritium fuel density ratio, however, is measurable in this setup only for trace levels of tritium, with an upper limit of nT/nD ≈ 6%, motivating further diagnostic exploration. Finally, modeling suggests that in order to adequately measure the self-heating profile, the neutron camera system will require energy and pulse-shape discrimination to suppress otherwise overwhelming fluxes of low energy neutrons and gamma radiation.

Neutron Detector Needs for ITER

For a long time, nuclear fusion has been anticipated to become a future power source. The International Thermonuclear Experimental Reactor (ITER) tokamak is designed to demonstrate the feasibility of applying the deuterium-tritium fusion reaction to human power needs. The measurements of ITER's fusion neutron flux parameters can provide information on total fusion power and fusion power density as well as other plasma parameters. This paper gives an overview of the technical constraints in terms of the radiological, thermal, and electromagnetic loads for ITER neutron detectors. These constraints have been studied and summarized with measurement requirements. The areas of high risk have been highlighted to encourage research and development of neutron detectors for the urgent needs of ITER.

Dynamics of a reconnection-driven runaway ion tail in a reversed field pinch plasma

While reconnection-driven ion heating is common in laboratory and astrophysical plasmas, the underlying mechanisms for converting magnetic to kinetic energy remain not fully understood. Reversed field pinch discharges are often characterized by rapid ion heating during impulsive reconnection, generating an ion distribution with an enhanced bulk temperature, mainly perpendicular to magnetic field. In the Madison Symmetric Torus, a subset of discharges with the strongest reconnection events develop a very anisotropic, high energy tail parallel to magnetic field in addition to bulk perpendicular heating, which produces a fusion neutron flux orders of magnitude higher than that expected from a Maxwellian distribution. Here, we demonstrate that two factors in addition to a perpendicular bulk heating mechanism must be considered to explain this distribution. First, ion runaway can occur in the strong parallel-to-B electric field induced by a rapid equilibrium change triggered by reconnection-based relaxation; this effect is particularly strong on perpendicularly heated ions which experience a reduced frictional drag relative to bulk ions. Second, the confinement of ions varies dramatically as a function of velocity. Whereas thermal ions are governed by stochastic diffusion along tearing-altered field lines (and radial diffusion increases with parallel speed), sufficiently energetic ions are well confined, only weakly affected by a stochastic magnetic field. High energy ions traveling mainly in the direction of toroidal plasma current are nearly classically confined, while counter-propagating ions experience an intermediate confinement, greater than that of thermal ions but significantly less than classical expectations. The details of ion confinement tend to reinforce the asymmetric drive of the parallel electric field, resulting in a very asymmetric, anisotropic distribution.

Fusion proton diagnostic for the C-2 field reversed configuration

Measurements of the flux of fusion products from high temperature plasmas provide valuable insights into the ion energy distribution, as the fusion reaction rate is a very sensitive function of ion energy. In C-2, where field reversed configuration plasmas are formed by the collision of two compact toroids and partially sustained by high power neutral beam injection [M. Binderbauer et al., Phys. Rev. Lett. 105, 045003 (2010); M. Tuszewski et al., Phys. Rev. Lett. 108, 255008 (2012)], measurements of DD fusion neutron flux are used to diagnose ion temperature and study fast ion confinement and dynamics. In this paper, we will describe the development of a new 3 MeV proton detector that will complement existing neutron detectors. The detector is a large area (50 cm(2)), partially depleted, ion implanted silicon diode operated in a pulse counting regime. While the scintillator-based neutron detectors allow for high time resolution measurements (∼100 kHz), they have no spatial or energy resolution. The proton detector will provide 10 cm spatial resolution, allowing us to determine if the axial distribution of fast ions is consistent with classical fast ion theory or whether anomalous scattering mechanisms are active. We will describe in detail the diagnostic design and present initial data from a neutral beam test chamber.

Materials R&D for a timely DEMO: Key findings and recommendations of the EU Roadmap Materials Assessment Group

The findings of the EU Fusion Programme's ‘Materials Assessment Group’ (MAG), assessing readiness of Structural, Plasma Facing (PF) and High Heat Flux (HHF) materials for DEMO, are discussed. These are incorporated into the EU Fusion Power Roadmap [1], with a decision to construct DEMO in the early 2030s.The methodology uses project-based and systems-engineering approaches, the concept of Technology Readiness Levels, and considers lessons learned from Fission reactor material development. ‘Baseline’ materials are identified for each DEMO role, and the DEMO mission risks analysed from the known limitations, or unknown properties, associated with each baseline material. R&D programmes to address these risks are developed. The DEMO assessed has a phase I with a ‘starter blanket’: the blanket must withstand ≥2MWyrm−2 fusion neutron flux (equivalent to ∼20dpa front-wall steel damage). The baseline materials all have significant associated risks, so development of ‘Risk Mitigation Materials’ (RMM) is recommended. The R&D programme has parallel development of the baseline and RMM, up to ‘down-selection’ points to align with decisions on the DEMO blanket and divertor engineering definition. ITER licensing experience is used to refine the issues for materials nuclear testing, and arguments are developed to optimise scope of materials tests with fusion neutron (‘14MeV’) spectra before DEMO design finalisation. Some 14MeV testing is still essential, and the Roadmap requires deployment of a ≥30dpa (steels) testing capability by 2026. Programme optimisation by the pre-testing with fission neutrons on isotopically- or chemically-doped steels and with ion-beams is discussed along with the minimum 14MeV testing programme, and the key role which fundamental and mission-oriented modelling can play in orienting the research.

Fusion Nuclear Science Facility (FNSF) before Upgrade to Component Test Facility (CTF)

The compact (R0~1.2-1.3m) Fusion Nuclear Science Facility (FNSF) is aimed at providing a fully integrated, continuously driven fusion nuclear environment of copious fusion neutrons. This facility would be used to test, discover, and understand the complex challenges of fusion plasma material interactions, nuclear material interactions, tritium fuel management, and power extraction. Such a facility properly designed would provide, initially at the JET-level plasma pressure (~30%T2) and conditions (e.g., Hot-Ion H-Mode, Q<1)), an outboard fusion neutron flux of 0.25 MW/m2 while requiring a fusion power of ~19 MW. If and when this research is successful, its performance can be extended to 1 MW/m2 and ~76 MW by reaching for twice the JET plasma pressure and Q. High-safety factor q and moderate-plasmas are used to minimize or eliminate plasma-induced disruptions, to deliver reliably a neutron fluence of 1 MW-yr/m2 and a duty factor of 10% presently anticipated for the FNS research. Success of this research will depend on achieving time-efficient installation and replacement of all internal components using remote handling (RH). This in turn requires modular designs for the internal components, including the single-turn toroidal field coil center-post. These device goals would further dictate placement of support structures and vacuum weld seals behind the internal and shielding components. If these goals could be achieved, the FNSF would further provide a ready upgrade path to the Component Test Facility (CTF), which would aim to test, for ≤6 MW-yr/m2 and 30% duty cycle, the demanding fusion nuclear engineering and technologies for DEMO. This FNSF-CTF would thereby complement the ITER Program, and support and help mitigate the risks of an aggressive world fusion DEMO R&D Program. The key physics and technology research needed in the next decade to manage the potential risks of this FNSF are identified.

Fusion neutron flux of center hole in LiD assembly under CARR reactor

A LiD assembly was proposed to produce fusion neutrons from thermal neutrons in D2O region of the China Advanced Research Reactor (CARR). The transportation of produced fusion neutrons in the LiD assembly can obtain the fusion neutron flux in its central hole. The neutron flux and the distribution of 6Li(n,α) reaction rate in the LiD assembly were calculated by a computer simulation method with MCNP code. The calculation results show that the fusion neutron flux in the central hole of LiD assembly, Φ, is dependent on the size and structure of assembly, 6Li enrichment, tritium concentration in LiD rods and the physical state of LiD. In the case of LiD rods containing 5% tritium and 6Li enrichment 20%, Φ value reaches to 5.4×1013cm−2s−1, when the assembly was placed at the thermal neutron flux about 8.0×1014cm−2s−1 in CARR.

Neutron flux in lower hybrid current drive plasmas on EAST

Diagnostics application of fusion reaction measurements on the Experimental Advanced Superconducting Tokamak (EAST) has been proved by fusion neutrons during deuterium plasma discharges. In lower hybrid current drive plasmas, fusion neutrons are dominant mainly due to a sufficiently low loop voltage, which suppresses the photo-neutrons caused by runaway electrons. For fixed plasma current and lower hybrid wave (LHW) power, the fusion neutron flux depends on ion density as , stronger than a power of two, caused by an increased ion temperature, which is due to an increased electron and ion energy exchange by more frequent collisions at higher ion densities. The fusion neutron flux depends on , at fixed LHW power and main ion density, which is mainly due to ion temperature affected by ohmic power and energy confinement time. Finally, the unchanged fusion neutron flux with LHW power indicates little variation in the main ion temperature, which is consistent with direct ion temperature measurements. This is explained by the fact that an increased energy transfer to ions by collisions is almost compensated for by less frequent collisions and decreased energy confinement time at higher LHW powers.

Gas dynamic trap as high power 14 MeV neutron source

It is now widely recognized that further progress toward a commercial fusion reactor critically depends on the availability of low activated materials to be operated for many years in the fusion neutron environment without degradation of their properties. Therefore, high power neutron source (NS) for extensive material tests is in great demand. The realistic NS can be built on the basis of the gas dynamic trap (GDT) which is one of the simplest system for magnetic plasma confinement. GDT is an axisymmetric mirror machine of the Budker–Post type, but with a high mirror ratio ( R>10), and operated with warm and relatively dense plasma. Thus, due to frequent collisions, the plasma confined in a trap is very close to isotropic Maxwellian, and hence many instabilities are not excited and plasma behavior is similar to a classical one. It is proposed to use the oblique injection of neutral tritium and deuterium beams with an energy of the order of 100 keV into the GDT “warm” target plasma (electron temperature T e≥750 eV) to produce high intensity neutron flux. In this case, due to the ionization and charge-exchange collisions with plasma particles the energetic atoms will turn into ions confined in the trap. The axial density profile of these anisotropic energetic ions will be strongly inhomogeneous. Namely, the maximum of the fast ion density will be located in the vicinities of mirrors and the minimum—at the middle plane of the device. Thus, a strongly inhomogeneous fusion neutron flux, mostly generated in the fast deuterium–tritium (D–T) ion collisions will appear. GDT NS looks to be rather flexible system. It enables to construct the source stepwise, starting from rather moderate parameters. Increasing the testing zone length and neutral beam power one can increase both neutron flux and the flux density. Calculations show that even in the modest version of the NS with T e∼250 eV quite appreciable neutron flux density in the range of 0.2–0.3 MW/m 2 can be produced. At present, the maximum temperature about 130 eV is obtained in the experiments at the model GDT device at Budker INP. Further increase of the electron temperature in the device requires higher neutral beam power and now is considered to be straightforward.

Some experimental research on anisotropic effects in the neutron emission of dense plasma-focus devices

An axial plasma jet emitted from the dense plasma focus (DPF) was detected using a Faraday cup. The jet was registered into a cone with a half angle of . The angular measurements of the D - D fusion neutron flux, performed with CR-39 nuclear track detectors, allowed us to see that there is a high anisotropy of the neutron flux just in a narrow cone, practically coincident with that of the plasma jet. This phenomenon is consistent with a DPF thermonuclear model based on the hypothesis of axial inertial plasma confinement. The same Faraday cup measurements showed the existence of very collimated and brief (about 10 ns) ion beams. Image-converter pinch photographs showed that these beams could be produced by spatial charge-separation effects due to m = 0 instabilities. The beam energy spectrum, obtained with track detectors, had the range from 0.7 - 5 MeV. Both plasma jets and high energy beams can explain the high angular anisotropy of the neutron flux, in terms of non-thermal fusion mechanisms. On the basis of these experimental results we discuss the coexistence of thermal and non-thermal mechanisms for neutron production in a DPF.

Distributed sensing of fusion neutrons by plastic scintillating fibers

A flexible one-dimensionally distributed fast neutron sensor using plastic scintillating fibers (PSFs) is proposed for measuring neutron emission profile from plasmas, where the PSFs are set up around the fusion device and neutron flux distribution along the PSFs is measured by applying the time of flight (TOF) technique to the scintillated light propagation. From the performance test of this system with 5 m long PSF for 14 MeV neutrons, the counting efficiency was evaluated to be about 1.8 × 10 - s counts per incident neutron into the PSF core, and spatial resolution was reached up to about 15 cm (full width at half maximum). An experiment using a 14 MeV point neutron source was also made for demonstrating the feasibility of this system to measure neutron emission profile, where the PSF was set up in a circle of 60 cm diameter and the neutron distribution around a 14 MeV neutron source was measured by changing the position of the PSF circle relative to the neutron source. The results showed that the TOF spectrum along the PSF is sensitive enough to measure the neutron source position inside the circle.

Reclamation of tungsten from activated fusion reactor components

Tungsten or tungsten-rich alloys have been proposed as constructional materials for various near-plasma components in fusion reactors. In considering this application, two factors relating to the long-term availability of tungsten should be taken into account. Firstly, it should be noted that tungsten is a relatively rare element and has non-fusion applications for which it would be difficult to find substitutes and, secondly, that it exhibits a high rate of transmutation in a fusion neutron flux. Both factors indicate a need to conserved tungsten resources through the reclamation of used material. Re-use of the tungsten is favoured by the fact that its radioisotopes are relatively short-lived, hence chemical separation from its transmutation products after a cooling period of about 15 years would, ideally, yield a non-active recycled material. Several plausible processes are examined for segregation of tungsten from its principal transmutation products Os, Re, Ta, Ir, Pt and Hf, and a route based on chlorination followed by fractional distillation of the metal chlorides is suggested as the most promising.

Foil activation measurements of energy and flux of neutrons from D-T mixed beams driven into a target

Threshold foil activation analyses are presented for measuring the fusion neutron flux from D-T mixed beams driven into a solid target in a linear accelerator. Thin foils of Al, Ni, Cu, Zr, and Nb yielded fairly consistent neutron flux densities. The average energy of neutrons at the location of the foils was also determined by a method of cross section comparison, which showed good agreement with a theoretical prediction.

Feasibility of laser pumping with neutron fluxes from present-day large tokamaks

The minimum fusion-neutron flux needed to observe nuclear-pumped lasing with tokamaks can be reduced substantially by optimizing neutron scattering into the laser cell, located between adjacent toroidal-field coils. The laser lines most readily pumped are probably the 3He-Ne lines at 0.633 μm and in the infrared, where the 3He-Ne gas is excited by energetic ions produced in the 3He(n,p)T reaction. These lines are expected to lase at the levels of D-T neutron flux foreseen for the TFTR (Tokamak Fusion Test Reactor) in 1990 (≫1012 n/cm2/s), while amplification should be observable at the existing levels of D-D neutron flux (∼1010 n/cm2/s). Lasing on the 1.73 and 2.63 μm transitions of Xe may be feasible at the maximum expected levels of D-T neutron flux in TFTR enhanced by scattering.

Silicon surface barrier detector for fusion neutron spectroscopy

The usefulness of the intrinsic 28Si(n,α)25Mg and 28Si(n, p)28Al reactions in silicon surface barrier detectors for measurements of ion temperatures in D-T fusion plasmas and of D-T fusion neutron flux has been investigated. For a 14-MeV neutron-generator narrow-energy-line source, the energy resolution (∼1.5%), detection efficiency, and useful count rates have been determined. Based on these results, for a 100-ms time bin, D-T plasma ion temperatures from 2 to 6 keV can be determined from the full-width half-maximum (FWHM) of the neutron-induced α spectral lines, with the error estimated to be 25% to 5%, respectively. The maximum intrinsic detection efficiency for a nominal 1500-μm-thick detector is ∼6×10−4.

Transmutation and activation effects in high-conductivity copper alloys exposed to a first wall fusion neutron flux

Transmutation and activity characteristics are calculated for a number of high-conductivity copper-based alloys exposed to 2.5 y continuous irradiation in the first wall neutron flux of the Culham Conceptual Tokamak Reactor IIA with neutron power loading of 7 MW m −2. The computations are based on a modified form of the ORIGEN code and the cross section data library UKCTRIIIA. It is found that the copper base transmutes to other elements, principally nickel and zinc, at the rate of 0.28 wt% per MW y m −2. The probable effect of these unintended alloying additions on the thermal conductivity is briefly discussed. Since their activities are generally dominated by that of the copper component, the dilute alloys studied exhibit very similar activation and decay properties. The long term surface dose rate of alumina dispersion strengthened alloys may, however, be dominated by the γ decay of 26Al with half life 7.4 × 10 5 y. Comparison is made with the activation characteristics of type 316 austenitic steel and the martensitic steel HT-9. It is noted that the long-term activity of copper alloys may in practice be governed by their silver impurity content, unless this can be reduced to about 1 ppm.