Transport processes in an inertial confinement fusion reactor

The quasi one-dimensional method previously developed for calculating a transient, compressible, viscous flow through a complex array of tubes or jets was extended to include heat and mass exchange between the fluid and the jets. The application was again the impulsive crossflow of a lithium plasma through a close-packed annular arrangement of liquid jets, a problem that arises in the deisgn of inertial confinement fusion reactors. It was found that the peak hoop stress in the first wall of the reactor may derive from the direct impact of the plasma, rather than from the subsequent impact of the jets or fragments thereof. Depending on conditions in the cavity, the peak wall stress was calculated to be up to 4 times less when heat and mass transfer were accounted for. The sensitivity of the design to key parameters was established.     

Heat transfer in inertial confinement fusion reactor systems

The short time and deposition distance for the energy from inertial fusion products results in local peak power densities on the order of 10¹⁸ W/m³. This paper presents an overview of the various inertial fusion reactor designs which attempt to reduce these peak power intensities and describes the heat transfer considerations for each design.     

Heavy ion beam driven inertial confinement fusion target studies and reactor chamber neutronic analysis

A review of heavy ion beam driven target studies and reactor chamber neutronic analysis at the Institute for Neutron Physics and Reactor Technology of the Nuclear Research Center Karlsruhe is presented. Based on a single shell, multi-layered 4 mg deuterium-tritium (DT) target design, key issues of the inertial confinement fusion target behavior are discussed. These issues are: ion beam energy deposition within the target; equations of state for the different target materials under the extreme conditions encountered; numerical simulation of target compression, ignition and burn including possible instabilities; transport processes of photons, neutrons, gammas, and alphas; and reactor chamber neutronics including target spectra, blanket and first wall analysis. For most of these issues the present physics understanding is briefly reviewed, the available calculational tools are mentioned, and key problems, which need further research and development efforts, are identified. One important feature of this paper is a consistent treatment of the 4 mg DT target behavior and of the corresponding reactor chamber neutronic analysis.     

Target support for inertial confinement fusion

General Atomics (GA) plays an important industrial support role for the U.S. Inertial Confinement Fusion program in the area of target technology. This includes three major activities: target fabrication support, target handling systems development, and target chamber design. The work includes target fabrication for existing ICF experiments, target and target system development for future experiments, and target research and target chamber design for experiments on future machines, such as the National Ignition Facility (NIF).     

A 2-D model of plasma-jet interaction in the hylife inertial confinement fusion reactor

A two-dimensional, axisymmetric model was developed to study the impulsive crossflow of a lithium plasma through a close-packed annular arrangement of liquid jets, a problem that arises in the design of a certain class of inertial confinement fusion reactors. Calculations performed with this model indicate that a 5 m radius reactor cavity, standing 8 m high, can deliver 1 GWe when pulsed at 1 Hz, without adverse loading of the first wall.     

The pioneers' legacy of inertial confinement nuclear fusion

Some years ago Professor Guillermo Velarde visited his good friend Academician Nicolai G. Basov, Nobel Prize in Physics and Director of the Lebedev Institute, and discussed the possibility of publishing a book on the History of the Inertial Confinement Nuclear fusion to be written by its Pioneers. They thought that only the own Pioneers could focus the book on a realistic and truthful manner.The years passed and unfortunately on July 1, 2001, Nicolai G. Basov passed away. In June 2002 Professor Velarde was invited to Moscow to give a Memorial lecture on Academician Basov in the Academy of Sciences. There in Moscow, Professors Vladislav Rozanov, Sergey Guskov, Natividad Carpintero-Santamaría and Guillermo Velarde decided to carry on with the project.Unfortunately, it was time to close the deadlines and the chapter of Academician Robert Dautray, former Haut-Commissaire of the CEA was not received. We regretted very much that the contribution of this very relevant Pioneer to the ICF is not included in the book. We did not receive also any material from Professor Edouard Fabre from the Ecole Polytechnique.The absence of the contributions of Nicolai G. Basov and Edward Teller was very significant. All of us would like that this ICENES 2013, hosted by the Institute of Nuclear Fusion, could be a posthumous tribute to these two outstanding Pioneers as well as a tribute to all of them.     

惯性约束聚变实验中全程控终端同步机的研制

基于VXI总线系统设计了用于惯性约束聚变(ICF)实验的中等精度全程控终端同步机,可以对输出延时、幅度、极性、脉宽全面程控,并进行了性能指标测试.测试表明:该同步机固有延时小于30 ns,大信号输出脉冲前沿小于14 ns,时间抖动小于5 ns,满足了ICF实验对终端同步机的要求.     

Safety and environmental aspects of organic coolants for fusion facilities

Organic coolants, such as OS-84, offer unique advantages for fusion reactor applications. These advantages are with respect to both reactor operation and safety. The key operational advantage is a coolant that can provide high temperature (350–400°C) at modest pressure (2–4 MPa). These temperatures are needed for conditioning the plasma-facing components and, in reactors, for achieving high thermodynamic conversion efficiencies (>40%). The key safety advantage of organic coolants is the low vapor pressure, which significantly reduces the containment pressurization transient (relative to water) following a loss of coolant event. Also, from an occupational dose viewpoint, organic coolants significantly reduce corrosion and erosion inside the cooling system and consequently reduce the quantity of activation products deposited in cooling system equipment. On the negative side, organic coolants undergo both pyrolytic and radiolytic decomposition, and are flammable. While the decomposition rate can be minimized by coolant system design (by reducing coolant inventories exposed to neutron flux and to high temperatures), decomposition products are formed and these degrade the coolant properties. Both heavy compounds and light gases are produced from the decomposition process, and both must be removed to maintain adequate coolant properties. As these hydrocarbons may become tritiated by permeation, or activated through impurities, their disposal could create an environmental concern. Because of this potential waste disposal problem, consideration has been given to the recycling of both the light and heavy products, thereby reducing the quantity of waste to be disposed. Preliminary assessments made for various fusion reactor designs, including ITER, suggest that it is feasible to use organic coolants for several applications. These applications range from first wall and blanket coolant (the most demanding with respect to decomposition), to shield and vacuum vessel cooling, to an intermediate cooling loop removing heat from a liquid metal loop and transferring it to a steam generator or heat exchanger.     

Fatigue cracking of a bare steel first wall in an inertial confinement fusion chamber

Inertial confinement fusion power plants will deposit high energy X-rays onto the outer surfaces of the first wall many times a second for the lifetime of the plant. These X-rays create brief temperature spikes in the first few microns of the wall, which cause an associated highly compressive stress response on the surface of the material. The periodicity of this stress pulse is a concern due to the possibility of fatigue cracking of the wall. We have used finite element analyses to simulate the conditions present on the first wall in order to evaluate the driving force of crack propagation on fusion-facing surface cracks.Analysis results indicate that the X-ray induced plastic compressive stress creates a region of residual tension on the surface between pulses. This tension film will likely result in surface cracking upon repeated cycling. Additionally, the compressive pulse may induce plasticity ahead of the crack tip, leaving residual tension in its wake. However, the stress amplitude decreases dramatically for depths greater than 80–100 μm into the fusion-facing surface. Crack propagation models as well as stress-life estimates agree that even though small cracks may form on the surface of the wall, they are unlikely to propagate further than 100 μm without assistance from creep or grain erosion phenomena.     

Neutronics and photonics of inertial confinement fusion pellets with internal breeding

The neutronics and photonics performance of a pellet with a small DT core spark trigger, surrounded by a large volume of D to enable tritium and He-3 breeding, is examined. The response to a 70% DD and 30% DT composite neutron spectrum is calculated using either W, Be, or Pb as structural materials at core density radius products ranging from 9.42 to 94.2 kg/m². At a core density-product of 94.2, the DT neutron source leads to an excess particle multiplication of 0.43 neutrons per source neutron. The percentage of energy leakage from the pellet in the form of escaped neutrons is 42.3% of the source energy for the DT source, and 28.8% for the DD source. The gamma-ray energy percentage deposited in the pellet is 26.7% for the DT source and 106.6% for the DD source. For the pellet with the composite source, the energy multiplication factor is 1.27. Thus the large DD contribution to the composite neutron source results in the pellet performing many of the functions normally reserved for the blanket such as spectral softening, breeding, and neutron and energy multiplication. The neutron energy leakage is 38.4% of the source energy for the composite source. It is estimated that the neutron energy leakage amounts to 10% of the fusion energy, compared with 70% as neutron energy in a DT pellet. These results are significantly different from those encountered in conventional DT inertial confinement designs, and thus lower tritium inventories, higher power densities, reduced radiation damage, and materials activation of the reactor coolant and structure may be achievable.     

惯性约束聚变用聚酰亚胺靶丸的研究进展

综述了近年来国内外在惯性约束聚变实验用聚酰亚胺(PI)靶丸方面的研究与应用进展情况.从ICF实验用靶的性能要求、常用PI的化学结构、PI靶丸的装配方法以及目前PI靶丸应用过程中存在的问题等几个方面进行了论述.在此基础上提出了PI靶丸的研究方向.     

ICF用靶丸内燃料充填参数对温度的依赖

通过BWR(Bcncdict-Wcbb-Rubin)方程研究了内径分别为100、200、400、520μm的薄壁玻璃微球在500、400、300、77、35K的温度下,其内部的压力值和在工作温度(22K)下内部液层厚度之间的对应关系。结果发现,在低温下燃料的压力比常温下低很多。     

惯性约束聚变中成像的坐标关联技术

研究处理惯性约束聚变（ICF）中三维成像的各记录面间的关联定位问题。首先说明在ICF中采用三维成像的必要性,然后对适用于ICF的重建方法进行简要的描述,提出了通过光强重心进行坐标关联定位方法,定 位精度可达到10μm,彻底解决了实验中的针孔相机需要精确关联定位的问题。