Inertial fusion energy development beyond 2000

Conclusions Ultimately, of course, a prototype power plant will be built at a power level appropriate for planned future commercial operations. This could use the same ETF/ DPP driver or a new one tailored to the plant size and with less experimental flexibility than the ETF driver. With the experience and data gained from a number of small demonstration reactors, and from the operation of the ETF/DPP driver and target factory, it is quite likely that a variety of plant sizes options will be available at that time.The scenario explored here is a relatively low-cost development program for fusion energy, which encourages technology transfer to American industry at an early stage. If the government builds an ETF driver, target factory, a single-shot experiment area, and a burst mode facility, commercial companies may be interested in building their own small demonstration reactors which would be supported by the government facilities. The fact that the ETF and any number of DPPs could be supported by the same driver and target factory means that the incremental cost of trying many alternatives is small. The fact that IFE demonstration reactors can test all relevant parameters at low power means that IFE has no extremely high-cost (multi-billion dollar) development facility to build in order to demonstrate engineering feasibility, i.e., there is no large development hurdle to surmount. We can, indeed, start small and work our way larger as the results justify. The result of this approach may produce competitive IFE power plant designs from a few to a few thousand megawatts.     

Liquid lithium divertor system for fusion reactor

One of the most critical issues for the steady state fusion reactor is the heat flux in the divertor target. This paper proposes a liquid lithium divertor system to solve this problem. The proposed divertor system consists of a liquid lithium target, an evaporation chamber and a differential evacuation chamber. The heat coming from the fusion plasma along the divertor leg is removed by evaporation of lithium. The lithium vapor is condensed on the wall and is circulated with a pump. The coolant temperature for the wall is high enough to drive a power generator. Narrow slits along the divertor leg and the differential evacuation chamber reduce leakage of lithium vapor to the plasma chamber. A preliminary estimation predicts that the lithium ion density in the core plasma is lower than the plasma density.     

Safety Assessment for Inertial Fusion Energy Power Plants: Methodology and Application to the Analysis of the HYLIFE-II and SOMBRERO Conceptual Designs

Although the safety and environmental (S & E) characteristics of fusion energy have long been emphasized, these benefits are not automatically achieved. To maximize the potential S & E attractiveness of the inertial fusion energy (IFE), analyses must be performed early in the designs so that lessons can be learned and intelligent decisions made. In this work we have introduced for the first time heat transfer and thermal-hydraulics calculations as part of a state-of-the-art set of codes and libraries in order to establish an updated methodology for IFE safety analysis. We have focused our efforts primarily on two IFE power plant conceptual designs: HYLIFE-II and SOMBRERO. To some degree, these designs represent the extremes in IFE power plant designs. Also, a preliminary safety assessment has been performed for a generic target fabrication facility producing various types of targets and using various production techniques. Although this study cannot address all issues and hazards posed by an IFE power plant, it advances our understanding of radiological safety of such facilities. This will enable better comparisons between IFE designs and competing technologies from the safety point of view.     

Laser coupling of the numerous flattened Gaussian beams in the fiber-based ICF laser driver

Fiber-based inertial confinement fusion (ICF) laser driver provides a new pathway to realize the inertial fusion energy (IFE). The feasibility of this proposal is checked from the perspective of laser coupling process in this paper. Flattened Gaussian beam (FGB) is assumed for theoretical analysis. The focusing properties of the FGB are used to obtain the requirements for a single laser beam. Based on the typical parameters of the chamber and target in ICF research, the output energy from a single fiber amplification chain is estimated to be over several hundred milli-joule. New fiber structures needs to be designed to meet the requirements.     

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.     

An overview of the development of the first wall and other principal components of a laser fusion power plant

This paper introduces the JNM Special Issue on the development of a first wall for the reaction chamber in a laser fusion power plant. In this approach to fusion energy a spherical target is injected into a large chamber and heated to fusion burn by an array of lasers. The target emissions are absorbed by the wall and encapsulating blanket, and the resulting heat converted into electricity. The bulk of the energy deposited in the first wall is in the form of X-rays (1.0-100 keV) and ions (0.1-4 MeV). In order to have a practical power plant, the first wall must be resistant to these emissions and suffer virtually no erosion on each shot. A wall candidate based on tungsten armor bonded to a low activation ferritic steel substrate has been chosen as the initial system to be studied. The choice was based on the vast experience with these materials in a nuclear environment and the ability to address most of the key remaining issues with existing facilities. This overview paper is divided into three parts. The first part summarizes the current state of the development of laser fusion energy. The second part introduces the tungsten armored ferritic steel concept, the three critical development issues (thermo-mechanical fatigue, helium retention, and bonding) and the research to address them. Based on progress to date the latter two appear to be resolvable, but the former remains a challenge. Complete details are presented in the companion papers in this JNM Special Issue. The third part discusses other factors that must be considered in the design of the first wall, including compatibility with blanket concepts, radiological concerns, and structural considerations.     

Complexity And Availability for Fusion Power Plants: The Potential Advantages of Inertial Fusion Energy

Probably the single largest advantage of the inertial route to fusion energy (IFE) is the perception that its power plant embodiments could achieve acceptable capacity factors. This is a result of its relative simplicity, the decoupling of the driver and reactor chamber, and the potential to employ thick liquid walls. We examine these issues in terms of the complexity, reliability, maintainability and, therefore, availability of both magnetic and inertial fusion power plants and compare these factors with corresponding scheduled and unscheduled outage data from present day fission experience. We stress that, given the simple nature of a fission core, the vast majority of unplanned outages in fission plants are due to failures outside the reactor vessel itself. Given we must be prepared for similar outages in the analogous plant external to a fusion power core, this puts severe demands on the reliability required of the fusion core itself. We indicate that such requirements can probably be met for IFE plants. We recommend that this advantage be promoted by performing a quantitative reliability and availability study for a representative IFE power plant and suggest that databases are probably adequate for this task.     

Study on (n,t) Reactions of Zr, Nb and Ta Nuclei

The world faces serious energy shortages in the near future. To meet the world energy demand, the nuclear fusion with safety, environmentally acceptability and economic is the best suited. Fusion is attractive as an energy source because of the virtually inexhaustible supply of fuel, the promise of minimal adverse environmental impact, and its inherent safety. Fusion will not produce CO₂ or SO₂ and thus will not contribute to global warming or acid rain. Furthermore, there are not radioactive nuclear waste problems in the fusion reactors. Although there have been significant research and development studies on the inertial and magnetic fusion reactor technology, there is still a long way to go to penetrate commercial fusion reactors to the energy market. Because, tritium self-sufficiency must be maintained for a commercial power plant. For self-sustaining (D-T) fusion driver tritium breeding ratio should be greater than 1.05. And also, the success of fusion power system is dependent on performance of the first wall, blanket or divertor systems. So, the performance of structural materials for fusion power systems, understanding nuclear properties systematic and working out of (n,t) reaction cross sections are very important. Zirconium (Zr), Niobium (Nb) and Tantal (Ta) containing alloys are important structural materials for fusion reactors, accelerator-driven systems, and many other fields. In this study, (n,t) reactions for some structural fusion materials such as ^{88,90,92,94,96}Zr, ^93,94,95Nb and ^179,181Ta have been investigated. The calculated results are discussed andcompared with the experimental data taken from the literature.     

Safety Analysis of ITER Test Blanket Module for Water Cooled Blanket with Pebble Bed Breeder

Performance test of test blanket modules in the fusion environment using the International Thermonuclear Experimental Reactor (ITER) is one of the most important mile-stone for the development of the breeding blanket of the fusion power plant. In the design of test blanket modules in the ITER, it is very important to show that test modules do not cause additional safety concern to the ITER. This work has been performed for the evaluation of the preliminary safety of the test blanket module of a water cooled solid blanket, which is the primary candidate of the breeding blanket in Japan currently. Major issues of the evaluation were, establishment of post-accident cooling in the test blanket module, hydrogen gas generation by Be/steam reaction, and pressure increase and spilled water amount by the event of coolant leakage. The analyses results showed that, suppression tank system is necessary to accommodate the over-pressure by the coolant water after pipe break in the box of the test module. Coolant water pipe break of the first wall of the test blanket module will result relatively small impact to the ITER safety because of the small inventory of the coolant water of the test module and large volume of the vacuum vessel of the ITER. However, it was clarified that the water cooled blanket with beryllium pebble as the multiplier will have the potential hazard of the hydrogen formation. Further investigation to maintain the safety on this aspect is required.     

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.     

Threats, design limits and design windows for laser IFE dry wall chambers

The High Average Power Laser (HAPL) program is carrying out a coordinated effort to develop inertial fusion energy based on lasers, direct-drive targets and a dry wall chamber. The dry wall must accommodate the ion and photon threat spectra from the fusion micro-explosion over its required lifetime. This paper summarizes the current HAPL strategy on the armor/first wall configuration based on tungsten and ferritic steel as preferred armor and structural materials, respectively. The thermal performance of an example fully dense tungsten armor configuration on a ferritic steel first wall is described showing the basis for separating the high energy accommodation function of the armor from the structural function of the first wall. Example design operating windows for the armor, first wall and blanket are presented based on different requirements and constraints. The possibility of utilizing an engineered porous armor is discussed. Key chamber wall and armor issues are summarized.     

水冷包层模块第一壁流动传热特性初步分析

本文基于我国聚变工程实验堆水冷包层优化设计与安全分析的要求,针对水冷包层模块第一壁的流动传热特性进行三维数值模拟研究。采用计算流体力学方法,建立了水冷包层模块第一壁的三维数值模型,研究流量分配的特点以及温度分布情况,分析与评估在稳态工况、瞬态工况及失流事故下的水冷包层模块第一壁传热能力。研究结果表明,不同冷却管间存在流量分配不均匀的现象;在稳态工况下,水冷包层模块第一壁具有较好的传热能力,瞬态工况下水冷包层模块能够有效地导出反应堆热量;失流事故下冷却管内温度短时间上升至系统压力下的饱和温度,有待进一步研究。相关研究为优化包层第一壁传热设计提供参考,并为今后聚变堆的安全分析提供依据。     

Safety study of primary coolants in fusion power reactor first wall design

This paper examines potential safety problems associated with the various primary coolant candidates currently considered for the EPR fusion blanket designs. The basic concern is the possibility of overheating and melting of the first wall and the blanket, induced by a malfunction in the primary coolant system. These accidents include the loss-of-coolant flow, the loss-of-heat removal, overpower transients, and the loss of coolant. Following a mechanistic safety for these four types of accident sequences and comparing helium and liquid metal cooling, it was found that helium has a more adverse effect on the first-wall heat up in the event of a loss-of-heat removal or a loss-of-coolant because its lack of thermal inertia.     

Integrated inertial fusion energy chamber dynamics and response

This paper presents results of three-dimensional hydrodynamics simulations of the flow inside a model inertial fusion energy (IFE) fusion chamber. Turbulence modeling employing the large-eddy simulation approach is used to estimate the gas dynamics, state, and mixing after a sufficiently large number of target ignitions. The rich radiation-flow physics that takes place immediately after the lasers hit the hohlraum is modeled separately using a high-fidelity one-dimensional model, which provides reference conditions for the complex geometry three-dimensional turbulence simulations. The IFE geometry includes optical ports and recirculation openings as well as a duct to evacuate the debris produced after each energy deposition (as a model of a laser shot). Furthermore, a selected set of sensitivity studies are conducted to estimate the effect of uncertainty in radiative properties of the Xenon gas at the prevalent conditions in the chamber. The results provide guidance regarding the turbulence conditions in the chamber, which seem to have entered a decay state immediately before a new shot takes place. Computational estimates of the density variability within the chamber as well as pressure history at the approximate location of the laser optical ports is presented among other turbulence statistics.     

An Investigation for Ground State Features of Some Structural Fusion Materials

Environmental concerns associated with fossil fuels are creating increased interest in alternative non-fossil energy sources. Nuclear fusion can be one of the most attractive sources of energy from the viewpoint of safety and minimal environmental impact. When considered in all energy systems, the requirements for performance of structural materials in a fusion reactor first wall, blanket or diverter, are arguably more demanding or difficult than for other energy system. The development of fusion materials for the safety of fusion power systems and understanding nuclear properties is important. In this paper, ground state properties for some structural fusion materials as ²⁷Al, ⁵¹V, ⁵²Cr, ⁵⁵Mn, and ⁵⁶Fe are investigated using Skyrme–Hartree–Fock method. The obtained results have been discussed and compared with the available experimental data.     

Alternate approach to inertial confinement fusion with low tritium inventories and high power densities

A low-tritium-inventory, high-power-density, pool-type chamber approach to inertial confinement fusion is introduced. The concept uses target designs with internal tritium and³He breeding, eliminating the need for a lithium-breeding blanket. The fraction of the fusion energy carried out by neutrons is estimated as 10%, compared with 70% in a typical D-T system, and the neutron spectrum is softer. Liquid metals other than lithium that are less chemically reactive, such as lead, can be used for first-wall protection. The reduced neutron component and the elimination of the need for a thick lithium blanket for tritium breeding lead to higher power densities and more compact chamber designs. The radiation damage at the first structural wall is reduced, leading to potentially longer wall lifetimes. A significant environmental advantage in terms of reduced radioactive release risks under operational and accident conditions is identified, primarily due to the one to two orders of magnitude reduction in the tritium inventories compared with D-T-based systems.     

水冷固态增殖包层模块冷却剂管系流量分配研究

包层是磁约束聚变堆中实现氚增殖和能量导出的重要部件,针对包层模块中,由于复杂的串并联流道结构所导致的冷却剂流量分配不均匀问题,采用一维热流体流动分析软件Flowmaster,建立了水冷固态增殖包层子模块的冷却剂流道结构模型。对运行工况下包层冷却剂流量分配进行模拟,并与相关试验以及模拟结果进行比对。模拟结果表明,所建立的子模块一维模型各部分冷却剂温升和压降均与设计值吻合,模型能够准确的描述包层冷却剂流动特性。在稳态运行工况下,包层子模块侧壁支管出现较为明显的流量分配不均匀现象,流量最大值与最小值偏差达到5%。位于侧壁上下两端的集合管对流量分配均匀性起重要作用,保持矩形集合管横截面积不变,横截面长宽相等时流量分配最为均匀。当集合管采用不同形状设计时,圆形管道流量分配均匀性要好于矩形管道。     

双冷嬗变包层第一壁热工水力特性数值模拟研究

利用CFX程序对聚变驱动次临界堆嬗变包层第一壁在水冷条件下的热工水力特性进行数值模拟和分析。计算选用PWR典型工况下的水,取嬗变包层第一壁的局部模型,考虑了流固热耦合,重点计算分析了在不同壁面热流密度和冷却剂流速条件下冷却剂温度、压降及结构材料最高温度的分布。计算结果显示,当水的入口流速为1～5m/s时,结构材料的最高温度较使用典型工况下的氦气作冷却剂时低16～91K,同时结构材料最大温差降低了12.2%～49.5%。结果表明：水可较好地满足稳态工况下第一壁的换热要求。     

Reducing Effective Liquid Wall Thickness in a HYLIFE-II Fusion Breeder

One of the major inertial fusion energy reactor designs is HYLIFE-II which uses protective flowing liquid wall between fusion plasma and solid first wall. The most attractive aspect of this reactor is that protective liquid wall eliminates the frequent replacement of the first wall structure during reactor lifetime. Liquid wall thickness must be at least the thickness required for supplying sufficient tritium for the deuterium–tritium (DT) driver and satisfying radiation damage on the first wall below the limits. Reducing this thickness results less pumping power requirements and cost of electricity. In this study, investigation on potential of utilizing refractory alloys (W-5Re, TZM and Nb-1Zr) as first wall to reduce effective liquid wall thickness in HYLIFE-II reactor using liquid wall of Flibe + 10 mol % UF₄ mixture. Neutron transport calculations were carried out with the help of the SCALE4.3 system by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and a S₈-P₃ approximation. Numerical results showed that using W-5Re or TZM as first wall was effective in decreasing liquid wall thickness in contrast to Nb-1Zr.     

Thermomechanical processes in an inertial thermonuclear fusion reactor blanket under cyclic exposure to neutron fluence

The thermomechanical processes occurring in the blanket of a heavy-ion inertial thermonuclear fusion reactor are analyzed. The heat-release density in the structural materials and the blanket coolant is calculated on the basis of prescribed characteristics of the neutron pulse from a ~1 GJ microexplosion of a target. Calculations of the nonstationary thermoelastic stress and pressure fields as well the temperature of the wall and blanket coolant are performed using simple one-dimensional models. The amplitude and frequency characteristics of the stress and pressure waves generated by the energy released in the microexplosion are obtained. The thermal relaxation processes occurring in the elements of the blanket between successive microexplosions are analyzed.