The Potential of Fast Ignition and Related Experiments with a Petawatt Laser Facility

A model of energy gain induced by fast ignition of thermonuclear burn in compressed deuterium-tritium fuel, is used to show the potential for 300× gain with a driver energy of 1 MJ, if the National Ignition Facility (NIF) were to be adapted for fast ignition. The physics of fast ignition has been studied using a petawatt laser facility at the Lawrence Livermore National Laboratory. Laser plasma interaction in a preformed plasma on a solid target leads to relativistic self-focusing evidenced by x-ray images. Absorption of the laser radiation transfers energy to an intense source of relativistic electrons. Good conversion efficiency into a wide angular distribution is reported. Heating by the electrons in solid density CD₂ produces 0.5 to 1 keV temperature, inferred from the D-D thermo-nuclear neutron yield.     

Effect of Internal Breeding of Tritium and Helium-3 on the Ignition of an ICF Fuel Pellet

Self-heating condition and following ignition in an Inertial Confinement Fusion (ICF) fuel pellet is evaluated by calculating the power equations, dynamically. In fact, the self-heating condition is a criterion that determines the minimum parameters of a fuel (such as temperature, density and areal density) that can be ignited. Deuterium is the main component of ICF fuels as large amounts of it are naturally available. In addition, the use of deuterium as a fuel in ICF causes the production of tritium and helium-3. However, pure deuterium has a high ignition temperature (\(\hbox {T}\ge 40\,\hbox {keV}\)) which makes it inefficient. In this paper, the power equations are solved, dynamically, and it has been indicated that internal tritium and helium-3 production at early evolution of compressed deuterium fuel causes ignition at lower predicted temperatures.     

Application of High Energy Protons in Inertial Confinement Fusion by Using the Fast Ignitor Concept for DT Fusion

A short-laser-pulse driven ion flux is examined as a fast ignitor candidate for inertial confinement fusion. The main mechanism for ion acceleration is charge separation in a plasma due to high-energy electrons driven by the laser inside the target. Another very new branch of fast ignition research is the investigation of the use of laser generated proton beams. In the present paper aims to provide insights into the feasibility of the fast ignition concept with high energy beams of protons generated in laser–plasma interactions. The optimum parameters of an ion beam and laser pulse that are suitable for an ignition spark in a hot precompressed DT fuel are estimated as a rough guide. Also, in this paper we estimate the radius of Deuterium–Tritium (DT) fuel pellet that is equal to the protons range in DT plasma.     

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.     

Transient thermo-mechanical behavior of a direct-drive target during injection in an inertial fusion energy chamber

During injection in an inertial fusion energy (IFE) chamber, a direct-drive target is subject to heat loads from chamber wall radiation and energy exchange from the chamber gas constituents. These heat loads can lead to the deuterium-tritium (DT) reaching its triple point temperature and even undergoing phase change, leading to unacceptable non-uniformity based on target physics requirements for compression and ignition of the DT fuel pellets using multiple laser beams. A two-dimensional bubble nucleation mode was added to the previously presented thermo-mechanical model to help better define the design margin for direct-drive IFE targets. The new model was validated by comparison with analytical results for controlled cases. It was then used to simulate heating experiments on DT targets conducted at the Los Alamos National Laboratory (LANL), where the ³He present in the DT due to tritium decay was found to affect the nucleation process.The previous requirement for target survival was for the temperature of the DT to remain below triple point of DT (19.79 K). If the existence of a melt layer does not violate the symmetry requirements on the target for successful implosion, the constraint could be relaxed by assuming a limit based on the avoidance of bubble nucleation. This study shows that the thresholds for melting and bubble nucleation are significantly different, allowing for extra margin in target survival under this assumption.     

High gain requirements and high field tokamak experiments

Operation at sufficiently high gain (ratio of fusion power to external heating power) is a fundamental requirement for tokamak power reactors. For typical reactor concepts, the gain is greater than 25. Self-heating from alpha particles in deuterium-tritium plasmas can greatly reducen/temperature requirements for high gain. A range of high gain operating conditions is possible with different values of alpha-particle efficiency (fraction of alpha-particle power that actually heats the plasma) and with different ratios of self heating to external heating. At one extreme, there is ignited operation, where all of the required plasma heating is provided by alpha particles and the alpha-particle efficiency is 100%. At the other extreme, there is the case of no heating contribution from alpha particles.n/temperature requirements for high gain are determined as a function of alpha-particle heating efficiency. Possibilities for high gain experiments in deuterium-tritium, deuterium, and hydrogen plasmas are discussed.     

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.     

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.     

A fusion–fission hybrid reactor with water-cooled pressure tube blanket for energy production

A fusion–fission hybrid reactor is proposed to achieve the energy gain of 3000 MW thermal power with self-sustaining tritium. The hybrid reactor is designed based on the plasma conditions and configurations of ITER, as well as the well-developed pressurized light water cooling technologies. For the sake of safety, the pressure tube bundles are employed to protect the first wall from the high pressure of coolant. The spent nuclear fuel discharged from 33GWD/tU Light Water Reactors (LWRs) and natural uranium oxide are taken as driver fuel for energy multiplication. According to thermo-mechanics calculation results, the first wall of 20 mm is safe. The radiation damage analysis indicates that the first wall has a lifetime of more than five years. Neutronics calculations show that the proposed hybrid reactor has high energy multiplication factor, tritium breeding ratio and power density; the fuel cannot reach the level of plutonium required for a nuclear weapon. Thermal-hydraulic analysis indicates that the temperatures of the fuel zone are well below the limited values and a large safety margin is provided.     

Visualizing an ignition process of hydrogen jets containing sodium mist by high-speed imaging

In severe accident scenarios for sodium-cooled fast reactors, it is desirable to gradually consume hydrogen generated by various ex-vessel phenomena without posting a challenge to containment integrity. An effective means is combustion of hydrogen jets containing sodium vapor and mist, but previous studies have been limited to determining ignition thresholds experimentally. The aim of this study was to visualize the ignition process in detail to investigate the ignition mechanism of hydrogen–sodium mixed jets. The ignition experiments of the hydrogen jet containing sodium mist were carried out under a condition of little turbulence. The ignition process was measured with an optical measurement system comprised of a high-speed camera and an image intensifier, and a spatial distribution of luminance was analyzed by image processing. Detail observation revealed that sodium mist particles burned as scattering sparks inside the jet and that hydrogen ignited around the mist particles. Additionally, the experimental results and a simple heat balance calculation indicated that the combustion heat of sodium mist particles could ignite the hydrogen as the heterogeneous ignition source in the fuel temperature range where the mist particle formation was promoted.     

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).     

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.     

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.     

On the safety of conceptual fusion-fission hybrid reactors

A preliminary examination of some potential safety questions for conceptual fusion-fission hybrid reactors is presented in this paper. The study and subsequent analysis was largely based upon one design, a conceptual mirror fusion-fission reactor, operating on the deuterium-tritium plasma fusion fuel cycle and the uranium-plutonium fission fuel cycle. The major potential hazards were found to be: (a) fission products, (b) actinide elements, (c) induced radioactivity, and (d) tritium.As a result of these studies, it appears that highly reliable and even redundant decay heat removal must be provided. Loss of the ability to remove decay heat results in melting of fuel, with ultimate release of fission products and actinides to the containment. In addition, the studies indicate that blankets can be designed which will remain subcritical under extensive changes in both composition and geometry. Magnet safety and the effects of magnetic fields on thermal parameters were also considered.     

Fission-suppressed blankets for fissile fuel breeding fusion reactors

Two blanket concepts for deuterium-tritium (DT) fusion reactors are presented which maximize fissile fuel production while at the same time suppress fission reactions. By suppressing fission reactions, the reactor will be less hazardous, and therefore easier to design, develop, and license. A fusion breeder operating a given nuclear power level can produce much more fissile fuel by suppressing fission reactions. The two blankets described use beryllium for neutron multiplication. One blanket uses two separate circulating molten salts: one salt for tritium breeding and the other salt for U-233 breeding. The other uses separate solid forms of lithium and thorium for breeding and helium for cooling.Nuclear power is the sum of fusion (D + T 14 MeV neutron+ 3.5 MeV alpha) power plus additional power from neutron-induced reactions in the blanket.     

聚变-裂变混合能源堆一维计算模型燃耗分析

聚变-裂变混合能源堆包括聚变中子源和以天然铀为燃料、水为冷却剂的次临界包层,主要目标是生产电力。利用输运燃耗耦合程序系统MCORGS计算了混合能源堆一维模型的燃耗,给出了中子有效增殖因数k_eff、能量放大倍数M、氚增殖比TBR等物理量随时间的变化。通过分析能谱和重要核素随燃耗时间的变化,说明混合能源堆与核燃料增殖、核废料嬗变混合堆的不同特点。本文给出的结果可作为混合堆中子输运、燃耗分析程序校验的参考数据,为混合堆概念研究提供了基础数据。     

氦气、水和熔盐冷却的混合堆中子学性能研究

氦气、水、熔盐(Flibe)在强磁场中流动不存在严重的MHD问题,因此适合在基于磁约束的聚变-裂变混合堆中作为冷却剂.针对氦气、水、Flibe这3种冷却剂对混合堆包层中子学性能的影响进行研究,分析包层中能谱特点及燃料增殖特性.通过燃耗计算,研究氚增殖率(TBR)、能量倍增因子(M)、keff等随运行时间的变化.中子学输运采用三维蒙特卡罗程序MCNP.计算结果表明,不同的冷却剂对混合堆系统中子能谱影响很大:氦冷系统的能谱最硬,主要发生快中子裂变,氚增殖效果最好;水冷系统的能谱最软,产能最多,但需提高TBR;Flibe冷系统的能谱较硬,产能最少.     

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.