DT Fusion Reactor Research Articles

Tritium permeation from tritiated water to water through Inconel

From a safety standpoint, it is important to understand the behavior of tritium in the coolant of DT fusion reactors. The high-temperature and high-pressure water is expected to be used as a coolant in JA DEMO. Therefore, it is necessary to deepen the understanding of the tritium transfer from the primary cooling water to the secondary cooling water in a heat exchanger. In this study, a double-tube permeation device consisting of Inconel 600 tube, which is a heat exchanger material, and SS316 tube, was assembled and the tritium permeation from tritiated water to water at 300 °C and 17 MPa was observed. Remarkable tritium permeation was observed after around 17 days as the cumulative heating time. Gaseous tritium (HT and T2) was detected in the gas phase of the tritiated water side after the permeation experiments. This suggests that a part of tritium generated in the metal oxidation reaction dissolved in the Inconel and diffused and permeated. With assuming that the diffusion process of tritium through the Inconel was the rate controlling step of permeation, the permeability of tritium through the Inconel tube from the tritiated water derived from the experimental data was 3 orders of magnitude smaller than that obtained from the hydrogen isotope permeation experiments from gas phase to gas phase.

Commissioning the new pulse dilation Gas Cherenkov Detector at the National Ignition Facility

The new Pulse Dilation – Photomultiplier Tube fielded on the Gas Cherenkov Detector GCD-3 at the National Ignition Facility allows for high band width DT gamma reaction history measurements. Crucial for successful data collection is precise timing of the dilation window. Using both a short pulse laser and test measurements on high yield ice layered DT shots the timing was set well within a 1-ns dilation window. This allowed for successful data collection on high yield shots for Carbon and Hohlraum/TMP (n,n’) gamma measurements and for DT fusion reaction rate measurements. The burn width measurements of GCD agree with burn width measurements of the sister instrument Gamma Reaction History.

Neutron backscatter edge: A measure of the hydrodynamic properties of the dense DT fuel at stagnation in ICF experiments

The kinematic lower bound for the single scattering of neutrons produced in deuterium-tritium (DT) fusion reactions produces a backscatter edge in the measured neutron spectrum. The energy spectrum of backscattered neutrons is dependent on the scattering ion velocity distribution. As the neutrons preferentially scatter in the densest regions of the capsule, the neutron backscatter edge presents a unique measurement of the hydrodynamic conditions in the dense DT fuel. It is shown that the spectral shape of the edge is determined by the scattering rate weighted fluid velocity and temperature of the dense DT fuel layer during neutron production. In order to fit the neutron spectrum, a model for the various backgrounds around the backscatter edge is developed and tested on synthetic data produced from hydrodynamic simulations of OMEGA implosions. It is determined that the analysis could be utilized on current inertial confinement fusion experiments in order to measure the dense fuel properties.

Development of a viable route for lithium-6 supply of DEMO and future fusion power plants

In the European DEMO program, the design development of a demonstration power plant (DEMO) is currently in its pre-conceptual phase. In DEMO, breeding blankets will use large quantities of lithium, enriched in the isotope lithium-6 (6Li), for breeding the tritium needed to feed the DT fusion reaction. Unfortunately, enriched lithium is commercially not available in the required quantities, which is threatening the success of future power plant applications of nuclear fusion. Even if the manufacturing of the breeding blankets is still two decades ahead of us, it is now mandatory to address the topic of lithium-6 supply and to make sure that a viable supply (and reprocessing) route is available when needed.This paper presents an unbiased systems engineering approach assessing a number of available lithium isotope separation methods by defining requirements, rating them systematically and finally calculating a ranking number expressing the value of different methods. As a result, we suggest using a chemical exchange method based on a lithium amalgam system, but including some important improvements leading to a more efficient and ‘clean’ process (the ICOMAX process) in comparison with the formerly used COLEX process. Furthermore, by modelling activities and experiments in the KIT mercury laboratory (HgLab Karlsruhe), it is shown which work has to be done in the next years to make sure that the technical-scale process is available in time to supply DEMO and future fusion power plants by middle of the 21st century.

Plasma kinetic effects on interfacial mix and burn rates in multispatial dimensions

The physics of mixing in plasmas is of fundamental importance to inertial confinement fusion and high energy density laboratory experiments. Two- and three-dimensional (2D and 3D) particle-in-cell simulations with a binary collision model are used to explore kinetic effects arising during the mixing of plasma media. The applicability of the one-dimensional (1D) ambipolarity condition is evaluated in 2D and 3D simulations of a plasma interface with a sinusoidal perturbation. The 1D ambipolarity condition is found to remain valid in 2D and 3D, as electrons and ions flow together required for J = 0. Simulations of perturbed interfaces show that diffusion-induced total pressure imbalance and hydroflows flatten fine interface structures and drive rapid atomic mix. The atomic mix rate from a structured interface is faster than the ∼t scaling obtained from 1D theory in the small-Knudsen-number limit. Plasma kinetic effects inhibit the growth of the Rayleigh-Taylor instability at small wavelengths and result in a nonmonotonic growth rate scaling with wavenumber k with a maximum at a low k value, much different from Agk (where A is the Atwood number and g is the gravitational constant) as expected in the absence of plasma kinetic effects. Simulations under plasma conditions relevant to MARBLE separated-reactant experiments on Omega and the NIF show kinetic modification of DT fusion reaction rates. With non-Maxwellian distributions and relative drifts between D and T ions, DT reactivity is higher than that inferred from rates using stationary Maxwellian distributions. Reactivity is also found to be reduced in the presence of finite-Knudsen-layer losses.

Conceptual design study for heat exhaust management in the ARC fusion pilot plant

The ARC pilot plant conceptual design study has been extended beyond its initial scope [B. N. Sorbom et al., FED 100 (2015) 378] to explore options for managing ∼525 MW of fusion power generated in a compact, high field (B0 = 9.2 T) tokamak that is approximately the size of JET (R0 = 3.3 m). Taking advantage of ARC’s novel design – demountable high temperature superconductor toroidal field (TF) magnets, poloidal magnetic field coils located inside the TF, and vacuum vessel (VV) immersed in molten salt FLiBe blanket – this follow-on study has identified innovative and potentially robust power exhaust management solutions. The superconducting poloidal field coil set has been reconfigured to produce double-null plasma equilibria with a long-leg X-point target divertor geometry. This design choice is motivated by recent modeling which indicates that such configurations enhance power handling and may attain a passively-stable detachment front that stays in the divertor leg over a wide power exhaust window. A modified VV accommodates the divertor legs while retaining the original core plasma volume and TF magnet size. The molten salt FLiBe blanket adequately shields all superconductors, functions as an efficient tritium breeder, and, with augmented forced flow loops, serves as an effective single-phase, low-pressure coolant for the divertor, VV, and breeding blanket. Advanced neutron transport calculations (MCNP) indicate a tritium breeding ratio of ∼1.08. The neutron damage rate (DPA/year) of the remote divertor targets is ∼3–30 times lower than that of the first wall. The entire VV (including divertor and first wall) can tolerate high damage rates since the demountable TF magnets allow the VV to be replaced every 1–2 years as a single unit, employing a vertical maintenance scheme. A tungsten swirl tube FLiBe coolant channel design, similar in geometry to that used by ITER, is considered for the divertor heat removal and shown capable of exhausting divertor heat flux levels of up to 12 MW/m2. Several novel, neutron tolerant diagnostics are explored for sensing power exhaust and for providing feedback control of divertor conditions over long time scales. These include measurement of Cherenkov radiation emitted in FLiBe to infer DT fusion reaction rate, measurement of divertor detachment front locations in the divertor legs with microwave interferometry, and monitoring “hotspots” on the divertor chamber walls via IR imaging through the FLiBe blanket.

A DT fusion reactor design in field-reversed configuration using normal conductive coils

Assuming continued stability, favorable energy confinement time scaling, and an effective current drive and maintenance methods, a feasible DT fusion reactor design is proposed for field-reversed configuration (FRC) which uses normal conductive copper magnetic field coils at room temperature. The reactor has 3 GW fusion power, 1.5 MW m−2 neutron wall loading at the first wall, and thermal loading less than 1 MW m−2 at diverter plates. Plasma has an almost straight cylindrical shape of 40 m in length and 8.6 m in diameter. FRC can obtain very high beta (over 0.7 in average) and the magnetic field strength of the reactor will be 1.215 T, which can be produced by normal conductive coils having 70 m in length, 17.6 m in diameter, and 1.5 m in thickness with 0.6 effective conductive area ratio. Its Ohmic power loss is ~74 MW, which is less than 10% of the expected electric power output. A scenario to reach ignition from the initial formation is considered. At first, two FRCs are formed at the both ends of the reactor by fast theta-pinch with a negative bias magnetic field 6 m in length and 0.5 m in diameter. The FRCs are accelerated up to 250 km s−1 by the gradient of magnetic field strength towards the center of the burning region, collide with each other, and form a single large FRC. Their kinetic energy is converted to thermal energy, and the merged FRC is 10 m in length and 1.8 m in diameter. This FRC plasma is brought to ignition by intensive neutral beam injection (NBI) heating and particle supply. Given 200 s heating duration, the maximum NBI power is ~250 MW before alpha particle heating becomes significant. After ignition, NBI heating is not required, but there is a possibility that some part of equilibrium current must be supplied by the NBI in the MeV region.

Necessary and sufficient conditions for practical fusion power

Editors’ note: We invited Steven Cowley, former CEO of the UK Atomic Energy Authority, to comment on points raised by Robert Hirsch.Cowley replies: Undoubtedly, tokamaks have yielded by far the best plasma confinement of all fusion experiments. Indeed, the Tokamak Fusion Test Reactor at the Princeton Plasma Physics Laboratory in New Jersey and the Joint European Torus (JET) at the Culham Centre for Fusion Energy in the UK have achieved stable fusion conditions and significant fusion power—up to 16 MW in JET—from the deuterium–tritium reaction. Furthermore, detailed modeling from models validated against experimental data predicts that the international tokamak experiment ITER will attain a fusion “burn,” a state in which external heating is negligible and self-heating by the fusion-generated alpha particles is sufficient or almost sufficient to sustain the discharge.A burn would be the long-awaited scientific demonstration that energy production from fusion is possible. Only ITER offers the chance of reaching that hugely important milestone in the next two decades. However, as Robert Hirsch indicates, ITER will not prove the economic viability of fusion power. Such a determination is nontrivial, and without further R&D it is necessarily uncertain.Hirsch is wrong that tokamak reactor studies have ended in most parts of the world. For example, at the time of writing, demonstration tokamak reactor designs are being developed in the European Union (EU),11. G. Federici et al., Fusion Eng. Des. 109–111, 1464 (2016). https://doi.org/10.1016/j.fusengdes.2015.11.050 South Korea,22. K. Kim et al., Nucl. Fusion 55, 053027 (2015). https://doi.org/10.1088/0029-5515/55/5/053027 and China,33. B. N. Wan et al., IEEE Trans. Plasma Sci. 42, 495 (2014). https://doi.org/10.1109/TPS.2013.2296939 and less directed reactor studies are being pursued by all other ITER partners. Those studies address the well-known and serious technical issues raised by Hirsch. The authors made no attempt to downplay their significance. To appreciate the depth of the analysis, one has to read the extensive literature. I can only summarize briefly the current understanding of each of Hirsch’s issues.In fission and in fusion, cost is determined by much more than the mass of the core. Detailed estimates of the cost of electricity from the 2006 EU fusion reactor designs put the range44. D. Maisonnier et al., Fusion Eng. Des. 81, 1123 (2006). https://doi.org/10.1016/j.fusengdes.2005.08.055 between 0.03 and 0.09 €/kWh. ITER’s cost overruns, which are expected to be significantly less than Hirsch’s estimate, reflect a project that requires extensive R&D at every stage. They do not reflect the intrinsic industrial cost of components. Nonetheless, it is important to understand the ITER costs much better. Recent research, such as on the suppression of plasma turbulence, and expected improvements in technology, such as for superconducting magnets,55. B. N. Sorbom et al., Fusion Eng. Des. 100, 378 (2015). https://doi.org/10.1016/j.fusengdes.2015.07.008 suggest that innovation will drive down the cost and scale of tokamak reactors. Although I would not take any cost estimates too seriously, they indicate that tokamaks may enter the market in the right cost range. It is simply too early to be conclusive about cost.Hirsch is correct in identifying the quenching of superconducting magnets as being an issue for nuclear regulators. In fact, it is an issue with the French nuclear regulator for ITER. Technical studies of ITER show that a rapid quench of the superconducting magnets, caused by impact or otherwise, would not breach the containment of the vacuum vessel, let alone the main containment of the cryostat. Thus such an accident, although costly, would not endanger the surrounding population.The radioactivity of DT fusion reactors is a well-known issue.44. D. Maisonnier et al., Fusion Eng. Des. 81, 1123 (2006). https://doi.org/10.1016/j.fusengdes.2005.08.055 Material scientists have developed low-activation steels that reduce key impurities—nickel, for example—so that the radio isotopes produced by neutron bombardment are short-lived. With such materials, the activated material made in a fusion power plant will be low-level waste after 100 years.Tokamak reactors also face challenges not mentioned by Hirsch: tritium breeding and storage, for example.Success is not assured, but it is far too early to say that tokamaks fail against the Electric Power Research Institute criteria. Stimulating innovation on a broader range of ideas is also desirable. But we have an opportunity with ITER to create a burning plasma with an output of approximately 500 MW of fusion power. That opportunity should not be missed.ReferencesSection:ChooseTop of pageReferences <<CITING ARTICLES1. G. Federici et al., Fusion Eng. Des. 109–111, 1464 (2016). https://doi.org/10.1016/j.fusengdes.2015.11.050, Google ScholarCrossref, ISI2. K. Kim et al., Nucl. Fusion 55, 053027 (2015). https://doi.org/10.1088/0029-5515/55/5/053027, Google ScholarCrossref, ISI3. B. N. Wan et al., IEEE Trans. Plasma Sci. 42, 495 (2014). https://doi.org/10.1109/TPS.2013.2296939, Google ScholarCrossref, ISI4. D. Maisonnier et al., Fusion Eng. Des. 81, 1123 (2006). https://doi.org/10.1016/j.fusengdes.2005.08.055, Google ScholarCrossref, ISI5. B. N. Sorbom et al., Fusion Eng. Des. 100, 378 (2015). https://doi.org/10.1016/j.fusengdes.2015.07.008, Google ScholarCrossref, ISI© 2017 American Institute of Physics.

Capture-gated Spectroscopic Measurements of Monoenergetic Neutrons with a Composite Scintillation Detector

We report on the measurements of monoenergetic neutrons from DD and DT fusion reactions by use of the capture gating method in a heterogeneous plastic-glass composite scintillation detector. The cylindrical detector is 5.08 cm in diameter and 5.05 cm in height and was fabricated using 1-mm diameter Li-doped glass rods (GS20) and scintillating polyvinyl toluene (EJ-290). Different scintillation decay constants are used to identify energy depositions in two materials constituting the composite scintillator. Geant4 simulations of the neutron thermalization and capture process were conducted, finding a mean capture time of approximately $2.6 \mu\hbox{s}$ for both DD and DT neutrons. A capture gating time acceptance window based on simulation results was used to identify the neutron thermalization pulses. The total scintillation light yield produced in neutron thermalization was measured and found to show consistency on event-by-event basis despite the variety of neutron thermalization histories prior to capture. The ratio of light yields from thermalization of 14.1 MeV and 2.45 MeV neutrons in the EJ-290 scintillator was determined to be 14.6, and the light output from 2.45 MeV neutrons was also correlated to its electron equivalent, obtaining a value of $0.58 \pm 0.05 \hbox{MeVee}$ .

Comparison of DT neutron production codes MCUNED, ENEA-JSI source subroutine and DDT

As the DT fusion reaction produces neutrons with energies significantly higher than in fission reactors, special fusion-relevant benchmark experiments are often performed using DT neutron generators. However, commonly used Monte Carlo particle transport codes such as MCNP or TRIPOLI cannot be directly used to analyze these experiments since they do not have the capabilities to model the production of DT neutrons. Three of the available approaches to model the DT neutron generator source are the MCUNED code, the ENEA-JSI DT source subroutine and the DDT code. The MCUNED code is an extension of the well-established and validated MCNPX Monte Carlo code. The ENEA-JSI source subroutine was originally prepared for the modelling of the FNG experiments using different versions of the MCNP code (−4, −5, −X) and was later extended to allow the modelling of both DT and DD neutron sources. The DDT code prepares the DT source definition file (SDEF card in MCNP) which can then be used in different versions of the MCNP code. In the paper the methods for the simulation of the DT neutron production used in the codes are briefly described and compared for the case of a simple accelerator-based DT neutron source.

Hydrodynamic growth and mix experiments at National Ignition Facility

Hydrodynamic growth and its effects on implosion performance and mix were studied at the National Ignition Facility (NIF). Spherical shells with pre-imposed 2D modulations were used to measure Rayleigh-Taylor (RT) instability growth in the acceleration phase of implosions using in-flight x-ray radiography. In addition, implosion performance and mix have been studied at peak compression using plastic shells filled with tritium gas and imbedding localized CD diagnostic layer in various locations in the ablator. Neutron yield and ion temperature of the DT fusion reactions were used as a measure of shell-gas mix, while neutron yield of the TT fusion reaction was used as a measure of implosion performance. The results have indicated that the low-mode hydrodynamic instabilities due to surface roughness were the primary culprits to yield degradation, with atomic ablator-gas mix playing a secondary role.

Near term, low cost, 14 MeV fusion neutron irradiation facility for testing the viability of fusion structural materials

For over 50 years, engineers have been looking for an irradiation facility that can provide a fusion reactor appropriate neutron spectrum over a significant volume to test fusion reactor materials that is relatively inexpensive and can be built in a minimum of time. The 14MeV neutron irradiation facility described here can nearly exactly duplicate the neutron spectrum typical of a DT fusion reactor first wall at damage rates of ≈4 displacements per atom and 40appm He generated over a 2l volume per full power year of operation. The projected cost of this multi-beam facility is estimated at ≈$20 million and it can be built in <4 years. A single-beam prototype, funded by the U.S. Department of Energy, is already being built to produce medical isotopes. The neutrons are produced by a 300keV deuterium beam accelerated into 4kPa (30Torr) tritium target. The total tritium inventory is <2g and <0.1g of T2 is consumed per year. The core technology proposed has already been fully demonstrated, and no new plasma physics or materials innovations will be required for the test facility to become operational.

Laser-driven high-energy-density deuterium and tritium ions for neutron production in a double-cone configuration

By using two-dimensional particle-in-cell simulations, we investigate laser-driven ion acceleration and compression from a thin DT foil in a double-cone configuration. By using two counterpropagating laser pulses, it is shown that a double-cone structure can effectively guide, focus, and strengthen the incident laser pulses, resulting in the enhanced acceleration and compression of D+ and T+. Due to the ion Coulomb repulsion and the effective screening from the external laser electric fields, the transverse diffusion of ions is significantly suppressed. Finally, the peak energy density of the compressed ions exceeds 2.73 × 1016 J/m3, which is about five orders of magnitude higher than the threshold for high energy density physics, 1011 J/m3. Under this condition, DT fusion reactions are initiated and the neutron production rate per volume is estimated to be as high as 7.473 × 1035/m3 s according to Monte Carlo simulations. It is much higher than that of the traditional large neutron sources, which may facilitate many potential applications.

Plane thermonuclear detonation waves initiated by proton beams and quasi-one-dimensional model of fast ignition

AbstractThe one-dimensional problem on bilatiral irradiation by proton beams of the plane layer of condensed DT mixture with length 2H and density ρ0 ≤ 100ρs, where ρs is the fuel solid-state density at atmospheric pressure and temperature of 4 K, is considered. The proton kinetic energy is 1 MeV, the beam intensity is 1019 W/cm2 and duration is 50 ps. A mathematical model is based on the one-fluid two-temperature hydrodynamics with a wide-range equation of state of the fuel, electron and ion heat conduction, DT fusion reaction kinetics, self-radiation of plasma and plasma heating by α-particles. If the ignition occurs, a plane detonation wave, which is adjacent to the front of the rarefaction wave, appears. Upon reflection of this detonation wave from the symmetry plane, the flow with the linear velocity profile along the spatial variable x and with a weak dependence of the thermodynamic functions of x occurs. An appropriate solution of the equations of hydrodynamics is found analytically up to an arbitrary constant, which can be chosen so that the analytical solution describes with good accuracy the numerical one. The gain with respect to the energy of neutrons G ≈ 200 at Hρ0 ≈ 1 g/cm2, and G &gt; 2000 at Hρ0 ≈ 5 g/cm2. To evaluate the ignition energy Eig of cylindrical targets, the quasi-1D model, limiting trajectories of α-particles by a cylinder of a given radius, is suggested. The model reproduces the known theoretical dependence Eig ~ ρ0−2 and gives Eig = 160 kJ for ρ0 = 100ρs ≈ 22 g/cm3.

Tritium management and anti-permeation strategies for three different breeding blanket options foreseen for the European Power Plant Physics and Technology Demonstration reactor study

In DT fusion reactors like DEMO, the commonly accepted tritium (T) losses through the steam generator (SG) shall not exceed about 2mg/d that are more than 5 orders of magnitude lower than the T production rate of about 360g/d in the breeding blanket (BB). A very effective mitigation strategy is required balancing the size and efficiency of the processes in the breeding and cooling loops, and the availability and efficiency of anti-permeation barriers. A numerical study is presented using the T permeation code FUS-TPC that computes all T flows and inventories considering the design and operation of the BB, the SG, and the T systems. Many scenarios are numerically analyzed for three breeding blankets concepts – helium cooled pebbles bed (HCPB), helium cooled lithium lead (HCLL), and water cooled lithium lead (WCLL) – varying the T processes throughput and efficiency, and the permeation regimes through the BB and SG to be either surface-limited or diffusion-limited with possible permeation reduction factor. For each BB concept, we discuss workable operation scenarios and suggest specific anti-permeation strategies.

Scrutiny on the Static Behavior of Physical Parameters for ITER90H-P Fusion Reactor Using Deuterium–Tritium and Deuterium-Helium Mixtures

In this work, the performance of fusion reactor ITER90H-P with considering D–T and D-3He fuels are examined by writing the dynamics equations on the system reactor. Therefore, we solve these equations analytically in the steady state. In this state we determine the optimum conditions for achieving the maximum fusion gain .In addition, we ignore the impurities because we need to high performance points without impurities. Our calculations in this paper show that we have maximum fusion gain for DT and D-3He fusion reactions in steady state at resonance temperature Kev70 for D-T fusion reaction respectively .Their maximum values of fusion gain are equal to 6.01 for D-T fuel-andthe0.012 for D-3He, respectively. Therefore, currently, using D-3He as a fusion fuel is not recommended.

Nucleation, growth and transport modelling of helium bubbles under nuclear irradiation in lead–lithium with the self-consistent nucleation theory and surface tension corrections

Helium (He) nucleation in liquid metal breeding blankets of a DT fusion reactor may have a significant impact regarding system design, safety and operation. Large He production rates are expected due to tritium (T) fuel self-sufficiency requirement, as both, He and T, are produced at the same rate. Low He solubility, local high concentrations, radiation damage and fluid discontinuities, among other phenomena, may yield the necessary conditions for He nucleation. Hence, He nucleation may have a significant impact on T inventory and may lower the T breeding ratio.A model based on the self-consistent nucleation theory (SCT) with a surface tension curvature correction model has been implemented in OpenFOAM® CFD code. A modification through a single parameter of the necessary nucleation condition is proposed in order to take into account all the nucleation triggering phenomena, specially radiation induced nucleation. Moreover, the kinetic growth model has been adapted so as to allow for the transition from a critical cluster to a macroscopic bubble with a diffusion growth process.Limitations and capabilities of the models are shown by means of zero-dimensional simulations and sensitivity analyses to key parameters under HCLL breeding unit conditions. Results provide a good qualitative insight into the helium nucleation phenomenon in LM systems for fusion technology and reinforces the idea that nucleation may not be a remote phenomenon, may have a large impact on the system's design and reveals the necessity to conduct experiments on He cavitation.

Specific equilibrium behavior of hydrogen isotopes adsorbed onto synthetic zeolite A-type governed by lithium cations

Since synthetic zeolites (SZs) are powerfully adsorptive for hydrogen isotopes at cryogenic temperatures such as liquefied nitrogen, adsorption processes using these have been considered applicable to such as recovery of tritium from the lithium blanket of DT fusion reactor system. Onto these zeolites the adsorptions isotherms for hydrogen isotopes onto SZ-NaA, SZ-CaA and SZ-NaX at 77.4K were already clarified experimentally and analytically. These isotherms exhibit similar profiles of Langmuir type. In this work, adsorption isotherms were examined for H2 and D2 on SZ-LiA at 77.4K. SZ-LiA was made from SZ-NaA by exchanging its sodium ions for lithium ones, provided by TOSOH Corp. The experimental results demonstrate the specific equilibrium behavior of hydrogen isotopes adsorbed on SZ-LiA, deviating from isothermal profiles on SZ-CaA and SZ-NaX. SZ-LiA show the isothermal profiles of adsorption for H2 and D2 similar to on the conventional zeolites in the range from around 1kPa to the atmospheric pressure, but exhibit a plateau around 1mol/kg between 0.1Pa and 100Pa, while other zeolites show linearly profiling isotherms. This deviation indicates the adsorptive power of SZ-LiA remarkably greater than that of the others.

Reaction analysis of neutron emission from D and DT plasmas with/without 3He

The effects of nuclear elastic scattering (knock-on) from energetic light ions on the neutron emission spectrum in fusion plasmas are studied. The knock-on spectral features in D and DT plasmas without and with radio frequency heating on a 3He minority component are determined through Monte Carlo simulations. The neutron production magnitudes are determined relative to those due to fuel ions in the thermal bulk state (down to the level of 10−5), and the interference of higher order terms in the outlying spectral regions where these dominate is evaluated. The study completes the mapping of nuclear reaction and elastic scattering effects on the neutron emission spectrum including first-, second- and third-order processes. Included are the cross-couplings between dd and dt fusion reactions as manifested in the full neutron spectrum (En = 0–20 MeV) for DT plasmas of varying isotopic composition. This study determines the relationship between components of the neutron emission spectra and those of the underlying ion velocity distributions, which provides a general basis for judging the plasma diagnostic information that can potentially be extracted from measurement and analysis of such neutron data.

Sensitivity of ICF ignition conditions to non-Maxwellian DT fusion reactivity

The hotspot ignition conditions in ICF are determined by considering the power balance between fusion energy deposition and energy loss terms. Uncertainty in any of these terms has potential to modify the ignition conditions, changing the optimum ignition capsule design. This paper considers the impact of changes to the DT fusion reaction rate due to non-thermal ion energy distributions. The DT fusion reactivity has been evaluated for a class of non-Maxwellian distributions representing a perturbation to the tail of a thermal distribution. The resulting reactivity has been used to determine hotspot ignition conditions as a function of the characteristic parameter of the modified distribution. The hotspot ignition conditions in ICF can be determined by considering the power balance between fusion alpha particle energy deposition, hydrodynamic work and energy loss via radiation and electron thermal conduction (1). The resulting conditions may be sensitive to uncertainties in any of the key plasma physics models or assumptions. This paper considers the impact of changes to the DT fusion reaction rate due to the presence of non-thermal ion energy distributions. The reactivity,� v� , for thermonuclear fusion reactions is conventionally derived by averaging cross- section data over a Maxwell-Boltzmann energy distribution. In many situations, such as when plasmas are strongly shocked or rapidly heated by energetic particles, reacting ions may be better characterised by a non-Maxwellian distribution. Since the self-relaxation time is longer for ions at higher energies (2), such effects typically produce an enhancement or depletion within the tail of the distribution. At temperatures where the reactivity is dominated by the contribution of particles in the tail, even small changes to the distribution may have a potentially significant impact on the fusion reaction rate. The DT fusion reactivity has been evaluated for a class of non-Maxwellian distributions, the q-Gaussian distribution (3), which represent a perturbation to the tail of a thermal distribution. The resulting reactivity has been used to calculate a revised ideal ignition temperature (4) and to determine hotspot ignition conditions as a function of the characteristic parameter of the distribution. This initial assessment does not consider in detail the physical development of non-Maxwellian distributions. If conditions are found to show significant sensitivity, then the generation of distributions may be explored in future work via kinetic simulation.