Investigation of Fuel Energy Gain for Tritium-Poor Fuels in Fast Ignition Fusion Approach

One of the most fascinating ignition schemes for the inertial fusion energy that might be feasible is fast ignition.Its targets are ignited on the outside surface so there is no need to low density and high temperature center is required by central hot spot ignition.Fast ignition concept is noteworthy for a simple but fundamental reason:In principle it requires less total energy input to achieve ignition.In this paper,fuel energy and fuel energy gain of nearly pure deuterium capsule are calculated.This capsule is ignited by a deuterium-tritium seed,which would reduce the tritium inventory to a few percentages.The variations of fuel energy gain versus fuel density have been studied and submitted.On the basis of different physical parameters the following results of the investigation are presented and discussed.The energy gain curves for different tritium concentrations are found and limiting gain curves are derived.Finally,tritium-poor fast ignitor is compared to equimolar deuterium-tritium fast ignitor.     

Impact of Solid Breeder Materials on Tritium Breeding in a Hybrid Reactor

Tritium breeding ratio (TBR) is one of the important parameters in design of a Deuterium–Tritium (DT) driven hybrid reactor. Therefore, selection of tritium breeder materials to be used in the blanket is very crucial. In this study, tritium breeding potential of the solid breeders, namely, or in a (DT) fusion driven hybrid reactor fuelled with or was investigated. For this purpose in addition to these solid breeders, different types of liquid breeders, namely natural lithium, Flibe, Flinabe and were used to examine the tritium breeding behavior of liquid–solid breeder couple combinations. Numerical calculations were carried out by using Scale 4.3. According to numerical results, the blanket with fuel using natural lithium as coolant and as solid breeder had the highest TBR value.     

Fusion of Multi-Pinched Plasma Beams Converging with Spatial Symmetry

Fusion reactions can be achieved by using deuterium from sea water as the fuel.The amount of deuterium in one gallon of sea water contains energy equivalent to three hundred gallons of gasoline.Satisfactory conditions of plasma temperature and density necessary to initiate fusion have been achieved in various research facilities.However,the confinement time is not sufficient for ignition due to plasma instabilities.Here we show that fatal plasma instabilities could be suppressed by the ingenious arrangement of multi-pinched plasma beams converging symmetrically in space based on the minimization principle of plasma potential energy.Confirmation tests are proposed using tiny wires containing deuterium.If successful,the results could lead to a feasible approach to obtaining commercial fusion power from sea water,hence without the need to use expensive and radioactive tritium as the fuel.     

Opacity Structure of Glass Ablator in ICF Target Design

Opacity is dependent on the radiation temperature, material temperature and density of the material. The collisional ionization and the excitation rate can have direct effect on the temperature and the electron density of plasma. The plasma density can be determined by measurements of Stark-broadened K-shell spectral lines. Because Silicon may be used as dopant in the ablator of ignition target, the knowledge of their opacities is very important. The purpose of this work is to obtain a more detailed structure of opacity in regards with broadening effects in the form of Voigt profile. For this aim the opacity frequency dependency and the mean opacity of mixed plasmas are calculated under local thermodynamic equilibrium conditions. The final results show that the Stark-broadened line shape is dependent on the density. Also it is shown that the resonance peak and spectrum broadening of opacity spectrum of mixed plasma such as the \(\hbox {SiO}_2\)-plasma is larger than a single atom plasma such as Silicon, Si. Eventually these results can be used in the design of fuel pellets in Inertial Confinement Fusion.     

Effect of tritium reduction in determining energy gain by using R-matrix method direct laser fusion in D-T reaction

The laser fusion criterion is known as the ρR-Criterion, also called high-gain condition. This parameter is temperature dependent and can be calculated by R-matrix method. This method is applied for determining improved fusion cross-section for the reactions T(d,n)4He, 3He(d,p)4He, D(d,p)T, D(d,n)3He. In this paper the time dependent reaction rate equations for fusion reaction T(d,n)4He are solved and by using the obtained results we computed the fu- sion power density, energy gain versus temperature and ρR-parameter. The obtained results show that a suitable com- bination may be a deuterium fraction fD=0.65 and fT=0.35 which would lead 30% reduction in the tritium content of the fuel mixture, and this choice would not change the energy gain value very much. Finally, the obtained energy gain for D-T reaction by using R-matrix is in good agreement with other theories.     

The Effect of Plasma Profiles on the Critical Value of $$n\uptau_{E}$$ for Ignition

The effect of plasma profiles for ignition condition in a stationary D–T plasma is investigated using the energy conservation equations for ions and electrons, assuming that steady state fusion power is produced with no external power. The alpha power heating is sufficiently large to sustain the plasma and to balance the combined Bremsstarhlung and thermal conduction losses. The space dependent Lawson criteria is derived and critical condition is identified. As a result of this analysis we have shown that the optimum temperature might be \(\bar{T} \approx 26\,{\text{keV}}\) and that the peaked profiles with \(n\sim\left( {1 - \frac{{r^{2} }}{{a^{2} }}} \right)^{{v_{n} }}\), ν _n  = 1, and \(T\sim\left( {1 - \frac{{r^{2} }}{{a^{2} }}} \right)^{{v_{T} }} ,\,v_{T} = 2\) are good to minimizing \(\bar{n}\uptau_{E}\) for ignition. The results for these profiles show the critical value of \((\bar{n}\uptau_{E} )_{min} = 0.08 \times 10^{20 } \,{\text{m}}^{ - 3} \,{\text{s}}\) showing the reduction by 1/3 from the reference value limit ν _n  = ν _T  = 0. For a 26 keV plasma with an energy confinement time of 1 s, a pressure of about 6.24 atm is required for the plasma to be ignited; that is, it is sustained purely by the self-heating of the fusion alpha particles.     

D-T experiments on TFTR

Temperatures, densities and confinement of deuterium plasmas confined in tokamaks have been achieved within the last decade that are approaching those required for a D-T reactor. As a result, the unique phenomena present in a D-T reactor plasma (D-T plasma confinement, alpha confinement, alpha heating and possible alpha driven instabilities) can now be studied in the laboratory. Recent experiments on the Tokamak Fusion Test Reactor (TFTR) have been the first magnetic fusion experiments to study plasmas with reactor fuel concentrations of tritium. The injection of 20 MW of tritium and 14 MW of deuterium neutral beams into the TFTR produced a plasma with a T/D density ratio of 1 and yielded a maximum fusion power of 9.2 MW. The fusion power density in the core of the plasma was 1.8 MW m^–3 approximating that expected in a D-T fusion reactor. In other experiments TFTR has produced 6.4 MJ of fusion energy in one pulse satisfying the original 1976 goal of producing 1 to 10 MJ of fusion energy per pulse. A TFTR plasma with T/D density ratio of 1 was found to have 20% higher energy confinement time than a comparable D plasma, indicating a confinement scaling with average ion mass, A, of _E. The core ion temperature increased from 30 keV to 37 keV due to a 35% improvement of ion thermal conductivity. Using the electron thermal conductivity from a comparable deuterium plasma, about 50% of the electron temperature increase from 9 keV to 10.6 keV can be attributed to electron heating by the alpha particles. At fusion power levels of 7.5 MW, fluctuations at the Toroidal Alfvén Eigenmode frequency were observed by the fluctuation diagnostics. However, no additional alpha loss due to the fluctuations was observed. These D-T experiments will continue over a broader range of parameters and higher power levels.Work supported by U.S. Department of Energy Contract No. DE-AC02-76-CHO-3073.     

Temperature distribution in mixed ThO2–UO2 fuel rods located in blanket of an inertial fusion energy breeder

Investigations of neutronic analysis and temperature distribution in fuel rods located in a blanket driven ICF (Inertial Confinement Fusion) have been performed for various mixed fuels and coolants under a first wall load of 5 MW/m². The fuel rods containing ThO₂ and UO₂ mixed by various mixing methods for achieving a flat fission power density are replaced in the blanket and cooled with different coolants; natural lithium, flibe, eutectic lithium and helium for the nuclear heat transfer. It is assumed that surface temperature of the fuel rod increases linearly from 500 °C (at top) to 700 °C (at bottom) during cooling fuel zone. Neutronic and temperature distribution calculations have been performed by MCNP4B Code and HEATING7, respectively. In the blanket fueled with pure UO₂ and cooled with helium, M (fusion energy multiplication ratio) increases to 3.9 due to uranium having higher fission cross-section than thorium. The high fission energy released in this blanket, therefore, causes proportionally increasing of temperature in the fuel rods to 823 °C. However, the M is 2.00 in the blanket fueled with pure ThO₂ and cooled with eutectic lithium because of more capture reaction than fission reaction. Maximum and minumum values of TBR (tritium breeding ratio) being one of main neutronic paremeters for a fusion reactor are 1.07 and 1.45 in the helium and the natural lithium coolant blanket, respectively. These consequences bring out that the investigated reactor can produce substantial electricity in situ during breeding fissile fuel and can be self-sufficient in the tritium required for the DT fusion driver in all cases of mixed fuels and coolant types. Quasi-constant fission power density profiles in FFB (fissile fuel breeding) zone are obtained by parabolically increasing mixture fraction of UO₂ in radial and axial directions for all coolant types. Such as, in the helium coolant blanket and the case of PMF (parabolically mixed fuel), Γ (peek-to-average fission power density ratio) of the blanket is reduced to 1.1, and the maximum temperatures of the fuel rods in radial direction of the FFB zone are also quasi-constant. At the same time, in the case of PMF, for all coolant types, the temperature profiles in the radial direction of the fuel rods rise proportionally with surface temperature from the top to the bottom of fuel rods in the axial direction. In other words, for each radial temperature profile in the axial direction, temperature differences between centerline and surface of the fuel rods are quasi-constant. According to the coolant types, these temperature diffences vary between 30 and 45 °C.     

Enhanced tritium production for fusion reactors via3He(n,p)3 H in the heavy water moderator of a CANDU reactor

There is an increasing requirement for tritium to supply the fuel needs of current experimental fusion devices and in the initial startup of future power generating reactors. Tritium is produced in heavy water reactors through deuterium activation, but the total production capacity of Canadian operated CANDUs will fall short of future demands, during the period before and for some time after self-sufficient reactors become available. Consequently, methods of enhancing tritium generating rates warrant investigation. Herein we provide the results of an inquiry into the feasibility of enhancing tritium production levels through the activation of helium-3 following its external addition to the heavy water moderator system of a hypothetical 500–600 MWe CANDU reactor. The approach adopted involves simulation of the temporal evolution of the tritium activities, originating from²H(n,)³H and³He(n, p)³H, as described by a simple first order kinetic model. The results suggest that the frequent addition of helium-3 to the moderator water will enhance tritium production inventories. The enhancement factor is highly dependent upon the rate at which helium-3 irretrievably escapes to the moderator cover gas. However, the direct activation of helium-3, contained in a closed loop such as the annulus gas system, for example, would be essentially complete within a few weeks without any significant loss.     

Candidates for Laser Fusion Energy with Minimized Radioactivity

The new candidates for laser fusion energy with minimized radioactivity were presented. The possibility of side-on laser ignition of H–¹¹B with negligible radioactivity encouraged to study the fusion of solid state H–⁷Li fuel which again turns out to be only about ten times more difficult than the side-on ignition of solid deuterium–tritium using petawatt-picosecond laser pulses at anomalous interaction conditions if very high contrast ratio. Updated cross sections of the nuclear reaction are included. In other words, the specific approach discussed here involves inducing a fusion burn wave without radioactivity by laser-driven impact of a relatively large block of plasma on the outside of a solid density H–¹¹B and H–⁷Li targets.     

Ignition of a Deuterium Micro-Detonation with a Gigavolt Super Marx Generator

The Centurion–Halite experiment demonstrated the feasibility of igniting a deuterium–tritium micro-explosion with an energy of not more than a few megajoule, and the Mike test, the feasibility of a pure deuterium explosion with an energy of more than 10⁶ MJ. In both cases the ignition energy was supplied by a fission bomb explosive. While an energy of a few megajoule, to be released in the time required of less than 10⁻⁹ s, can be supplied by lasers and intense particle beams, this is not enough to ignite a pure deuterium explosion. Because the deuterium–tritium reaction depends on the availability of lithium, the non-fission ignition of a pure deuterium fusion reaction would be highly desirable. It is shown that this goal can conceivably be reached with a “Super Marx Generator”, where a large number of “ordinary” Marx generators charge (magnetically insulated) fast high voltage capacitors of a second stage Marx generator, called a “Super Marx Generator”, ultimately reaching gigavolt potentials with an energy output in excess of 100 MJ. An intense 10⁷ Ampere-GeV proton beam drawn from a “Super Marx Generator” can ignite a deuterium thermonuclear detonation wave in a compressed deuterium cylinder, where the strong magnetic field of the proton beam entraps the charged fusion reaction products inside the cylinder. In solving the stand-off problem, the stiffness of a GeV proton beam permits to place the deuterium target at a comparatively large distance from the wall of a cavity confining the deuterium micro-explosion.

The Ignition Requirements of the Degeneracy Microspheres of Deuterium Helium-3 Mixture with Low-Radioactive

This paper examines the burn characteristics for inertial confinement D/³ He fuel pellets with different concentrations of Helium-3. It is shown that the Helium-3 relative density of the fuel mixture plays a significant role in determining the burn characteristics and fuel gain. In spite of the safety of the plasma degeneracy of D/³ He fuel with fraction of y?=?0.2 (y: Helium-3 content parameter), ignition of fuel is impossible. In design fuel extra to safety should be considered fractional burn-up and fuel gain. The main contribution of this research is to show that the plasma degeneracy of equimolar mixture of D/³ He fuel lowers the ignition temperature and increases fuel gain. The results indicate that a $n_{D} /n_{{^{3} He}}$ ?≤?0.3 is difficult to ignite reasonable driver energy. A fuel gain of 378 can be obtained with a D/³ He fuel with fraction of y?=?0.33, and areal density (ρR) of 12 g/cm². It is found that the fuel gain of an equimolar D/³ He fuel at temperature of 70?keV and ρR value of 8.5 g/cm² is 480. This value gain is higher by about 22% than the case of the pellets (y?=?0.33).     

Thermal Transport Effect in Tokamaks and Block Ignition for Laser Fusion

Measurements of thermal conduction in tokamaks parallel to the magnetic field were up to 20 times less than the classical values. This was explained by the quantum correction of the collision frequency of electrons with ions. This stowing effect of heat is applied to re-evaluate the ignition threshold for the energy flux density E* for the ignition of solid state density deuterium tritium using nonlinear (ponderomotive) laser force driven space charge neutral plasma blocks.     

Inertial Confinement Fusion with a Heavy-Ion Driver-Accelerator and a Cylindrical Target

An economically efficient power plant burning deuterium-tritium fuel can be built using a powerful heavy-ion accelerator of a new type. A multilinear cryogenic cylindrical target, 1 cm long, 0.44 cm in radius, and containing 7.8 mg equimolar DT fuel, is studied as an example. The driver-accelerator gives two different target irradiation regimes. In both regimes, the beams consist of platinum ions, accelerated up to 100 GeV, with different isotopic composition, charged in the first regime only positively and in the second regime positively and negatively. In the first regime, the beam energy is 4.8 MJ and the beam heats in 60 nsec only the target shell. High heating symmetry is achieved by rapidly rotating the beam around the target axis with frequency 1 GHz. The fuel is compressed into a dense filament, where the condition for propagation of a fusion burn wave is satisfied – R 0.4 g/cm². In the second regime, a beam with 0.4 MJ ions heats in 0.2 nsec compressed fuel with power density 2.5·10⁷ TW/cm² up to ignition temperature. The computed energy amplification factor in the target is 200.     

A Review of Confinement Requirements for Advanced Fuels

The energy confinement requirements for burning D-³He, D-D, or P-¹¹B are reviewed, with particular attention to the effects of helium ash accumulation. It is concluded that the DT cycle will lead to the more compact and economic fusion power reactor. The substantially less demanding requirements for ignition in DT (the n_{e E} T required for ignition in DT is smaller than that of the nearest advanced fuel, D-³He, by a factor of 50) will allow ignition, or significant fusion gain, in a smaller device; while the higher fusion power density (the fusion power density in DT is higher than that of D-³He by a factor of 100 at the same plasma pressure) allows for a more compact and economic device at fixed fusion power.     

Correcting raw diagnostic data for oscilloscope recording system distortions at the National Ignition Facility

The National Ignition Facility (NIF) is now producing experimental results for the study of inertial confinement fusion (ICF). These results are captured by complex diagnostic systems and are key to achieving NIF's goal to demonstrate thermonuclear burn of deuterium and tritium fuel in a laboratory setting. High bandwidth gamma-ray fusion-burn measurements and soft X-ray indirect and direct drive energetic measurements are both captured with oscilloscope recording systems that distort or modulate the raw data. The Shot Data Analysis team has developed signal processing corrections for these oscilloscope recording systems through an automated engine. Once these corrections are applied, accurate fundamental quantities can be discerned.     

The Essential Requirements of Transition to Non-equilibrium Burn Stage of DD Fuel in Simple Spherical Targets

In this research, the transition from equilibrium ignition to non-equilibrium burn is evaluated by calculating the energy balance equations analytically for targets which consist of inner DD fuel and surrounded by a high-Z pusher. It is expected that these targets can trap much of the produced charged particles, radiation or even fast neutrons because of their high-Z pusher. Accordingly, DD fuel can be ignited in volume ignition regime with low ignition temperatures of 35 keV compared to central ignition. Thus, to get a non-equilibrium burning stage, we have examined all the important gain and loss processes for these targets as the energy deposition of fusion products, thermal conduction, radiation flux, mechanical work, bremsstrahlung radiation and inverse Compton scattering as well as competition among them. These conditions have investigated for different areal densities of DD fuel in ρR?~?1–100 g/cm² and it is shown that as areal density rises, transition temperature decreases. But at high areal densities, the transition temperature does not vary significantly and the limiting temperature of ~?20 keV will be obtained. Also, transition into non-equilibrium burn is studied for such cases that thermonuclear burn occurs at stagnation moment, before and after that. It is observed that the positive and negative role of mechanical work on the transition conditions is very important and varies transition temperature remarkably. In all cases, transition temperature to non-equilibrium burn phase is always much lower than ideal ignition temperature in specific areal density.     

Saturation in deuterium retention of CFC graphite targets exposed to beryllium-seeded plasmas on PISCES-B

ITER strike-plates are foreseen to be of carbon-fiber-composite (CFC). In this study the CFC bulk deuterium retention in ITER-relevant conditions is investigated. DMS 701 (Dunlop) CFC targets were exposed to plasma in PISCES-B divertor plasma simulator. Samples were exposed to both pure deuterium plasma and beryllium-seeded plasma at high fluences (up to ) and high surface temperature (1070 K). The deuterium contents of the exposed samples have been measured using both thermal-desorption-spectrometry (TDS) during baking at 1400 K and ion beam nuclear reaction analysis (NRA). The total deuterium inventory has been obtained from TDS while NRA measured the deuterium depth distribution. In the analysed fluence range at target temperature of 1070 K, no fluence dependence was observed. The measured released deuterium is . In the case of target exposure with beryllium-seeded plasma no change in the released amount of deuterium was found. The deuterium concentration inside the samples is almost constant until the probed depth of ?m, except in the first 1 μm surface layer, where it is 5 times higher than in the bulk. No C erosion/redeposition was observed in the Be-seeded plasma cases. The measured retention, applied to 50 m² of ITER CFC surface, would imply a tritium saturated value of 0.3 gT, much lower than the ITER safety limit of 350 g.