Critical Science Issues for Direct Drive Inertial Fusion Energy

There are several topics that require resolution prior to the construction of an Inertial Fusion Energy [IFE] laboratory Engineering Test Facility [ETF]: a pellet that produces high gain; a pellet fabrication system that cost-effectively and rapidly manufactures these pellets; a sufficiently uniform and durable high repetition-rate laser pellet driver; a practical target injection system that provides accurate pellet aiming; and, a target chamber that will survive the debris and radiation of repeated high-gain pellet implosions. In this summary we describe the science issues and opportunities that are involved in the design of a successful high gain direct drive Inertial Confinement Fusion [ICF] pellet.     

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.     

The Rationale for an Expanded Inertial Fusion Energy Program

The rationale for an expanded effort on the development of inertial fusion as an energy source is discussed. It is argued that there should be a two-pronged, complementary approach to fusion energy development over the next two to three decades: (1) Magnetic Fusion (MFE) via ITER and the supporting magnetic domestic program and (2) Inertial Fusion (IFE), a credible, affordable approach that exploits unique US strengths and current world leadership. IFE is only a few years away from demonstration of single-shot ignition and fusion energy gain via NIF. Enhanced funding for IFE R&D is needed in the near-term in order to prepare to expeditiously proceed beyond NIF to the energy application of inertial fusion.     

Overview of the Preliminary Safety Analysis of the National Ignition Facility

The National Ignition Facility (NIF) is a proposed U.S. Department of Energy inertial confinement laser fusion facility. The candidate sites for locating the NIF are: Los Alamos National Laboratory, Sandia National Laboratory, New Mexico, the Nevada Test Site, and Lawrence Livermore National Laboratory (LLNL), the preferred site. The NIF will operate by focusing 192 individual laser beams onto a tiny deuterium-tritium target located at the center of a spherical target chamber. The NIF has been classified as a low hazard, radiological facility on the basis of a preliminary hazards analysis and according to the DOE methodology for facility classification. This requires that a safety analysis report be prepared under DOE Order 5481.1B, Safety Analysis and Review System. A Preliminary Safety Analysis Report (PSAR) has been approved, which documents and evaluates the safety issues associated with the construction, operation, and decommissioning of the NIF.     

Fusion safety and environmental R&D tasks in Japan

This paper aims at listing and evaluating the status of all the research and development (R&D) tasks necessary for the construction of a safe and environmentally benign fusion experimental reactor. At this time, it is not possible to define precisely the R&D tasks necessary for the licensing approval and those that are useful in improving safety but not necessarily required for licensing because the licensing procedure itself is still being discussed. Among the R&D tasks, the most important are considered to be those related to tritium safety, namely, those effective in reducing the uncertainty in tritium inventory in the plasma facing components and blanket, uncertainty in tritium permeation and leakage, and those to clarify tritium behavior in the containment and in the environment. The R&D tasks with second priority are judged to be those related to mobilization of the activation products such as activated erosion dust or the corrosion products. The volatilization of structural metal caused by the oxidation at high temperature seems to be highly unlikely but some experiments are needed to assure that this is the case.     

Diode-Pumped Solid-State Lasers for Inertial Fusion Energy

We have begun building the Mercury laser system as the first in a series of new generation diode-pumped solid-state lasers for inertial fusion research. Mercury will integrate three key technologies: diodes, crystals, and gas cooling, within a unique laser architecture that is scalable to kilojoule and megajoule energy levels for fusion energy applications. The primary near-term performance goals include 10% electrical efficiencies at 10 Hz and 100J with a 2–10 ns pulse length at 1.047 m wavelength. When completed, Mercury will allow rep-rated target experiments with multiple chambers for high energy density physics research.     

聚变能源中的氚化学与氚工艺研究进展及展望

核聚变被认为是人类社会未来的理想能源,对社会、经济的可持续发展具有重要的战略意义。氘氚聚变反应具有反应截面大、反应速率高、点火温度低及释放能量大等优点,是目前聚变研究的主要目标,而高效的氘氚燃料循环工艺与技术是实现聚变能源商业应用的基础。本文主要介绍氘氚燃料循环所涉及的等离子体排灰气中氚的快速回收、氚的增殖与提取、大规模氢同位素分离、氚测量等相关氚化学与氚工艺的研究进展及展望,以期对未来聚变能源氚工厂相关技术的研究提供借鉴和参考。     

我国聚变堆结构材料表面阻氚涂层的研究进展

阻氚涂层是聚变堆实现氚自持及氚安全的关键科学与技术问题之一。我国通过国家磁约束聚变能发展研究专项依托国内优势单位部署了阻氚涂层基础问题及工程化技术研发工作。本文介绍了国内外聚变堆结构材料表面阻氚涂层研究进展,重点评述了近几年我国在阻氚涂层的材料选择、制备技术及阻滞氢渗透机制三个科学技术问题的研究进展,提出今后的研究方向。目前我国阻氚涂层材料类型以氧化物涂层为主,涂层制备工艺技术在不断优化和更新。Al2O3/FeAl阻氚涂层的电化学沉积铝(ECA)、粉末包埋渗铝(PC)及热浸铝(HDA)等方法的工艺处理规模及涂层阻氚性能在国际上均相对领先。发展了研究阻氚涂层阻滞氢渗透作用机理的方法,将通常基于Fick定律的表象研究方法向原子级方法前推了一步。未来需在考虑涂层制备工艺与基体材料成分、性能的关系及其在复杂形状结构件的适用性基础上,开发长寿命、高阻氚性能的阻氚涂层材料及制备工艺。     

Safety and Environmental Activities for ITER

This paper will summarize highlights of the safety approach and discuss the ITER EDA safety activities. The ITER safety approach is driven by three major objectives: (1) Enhancement or improvement of fusion's intrinsic safety characteristics to the maximum extent feasible, which includes a minimization of the dependence on dedicated safety systems; (2) Selection of conservative design parameters and development of a robust design to accommodate uncertainties in plasma physics as well as the lack of operational experience and data; and (3) Integration of engineered mitigation systems to enhance the safety assurance against potentially hazardous inventories in the device by deploying well-established nuclear safety approaches and methodologies tailored as appropriate for ITER.     

Report of the FEAC Inertial Fusion Energy Review Panel: July 1996

This report presents the results and recommendations of the U. S. Department of Energy Fusion Energy Advisory Committee (FEAC) review of its Inertial Fusion Energy (IFE) program. The subpanel charged with the review was chaired by John Sheffield of Oak Ridge National Laboratory. The FEAC, to which the subpanel reported, was chaired by Robert Conn of the University of California at San Diego.     

Fusion Safety Studies in Russia in 1996

The paper seeks to provide a summary report of observations and results of some Russian fusion safety studies performed in 1996. Release of tritium and helium from neutron irradiated beryllium at relatively high neutron fluences has a burst nature. With the growth of the beryllium temperature-increase rate to 90 K/s, the temperature of tritium burst release decreases from 800 to 450–500°C and for helium decreases from 1200 to 500°C. Characterization of carbon and tungsten dust produced in experiments simulating plasma disruptions revealed that dust particle distribution of sizes for graphites and carbon fiber composites has a bimodal nature with maxima in the range of 0.01–0.03 and 2–4 m for composite UAM and in the range of 0.14–0.18 and 2–4 m for graphite MPG-8. Chemical reactivity of beryllium with air was studied as well. A mathematical model for beryllium weight gain under its chemical interaction with air at temperatures of 700–800°C as a function of beryllium porosity, temperature, and interaction duration was developed.     

Fusion technology for a magnetic fusion production reactor

The tandem mirror and tokamak are being considered as candidate fusion drivers for a materials production reactor that could be implemented in the 1990s. This report considers, in detail, the required performance characteristics of the fusion plasma and the major technological subsystems for each fusion driver. These performance characteristics are compared with the present state of the art, corresponding development needs are identified, and technology program requirements, in addition to those now being supported by the Department of Energy, are pointed out. The tandem mirror and tokamak fusion drivers are also compared with regard to their required advancements in plasma performance and technology development.This paper represents work carried out from 1980 to 1982 and was in draft form in 1982. It was received for publication with only minor editing of its 1982 version, explaining the fact that some of the material is dated.     

Condensed Atomic Hydrogen as a Possible Target in Inertial Confinement Fusion (ICF)

H atom Rydberg matter (RM) in excitation state n = 1 is concluded to be a form of metallic hydrogen [Badiei S, Holmlid L (2004) J Phys Condens Matter 16:7017]. This material can be produced at low pressure. This condensed form of hydrogen may be very useful as a dense hydrogen inertial confinement fusion (ICF) target, being almost metallic and ten times denser than solid (frozen) diatomic hydrogen used at present. Coulomb explosions and plasma formation are initiated in condensed atomic hydrogen even by relatively weak nanosecond pulsed lasers. The protons emitted with high directivity in these explosions are energetic, corresponding to T = 10⁵ K, and they may be utilized to give strong compression of the material. The fastest protons observed at up to 1 keV indicate a compression considerably higher than that required for “fast ignition” fusion.     

Doses from accidental releases of tritium and activation products into the atmosphere

In view of public acceptance and the licensing procedure of projected fusion reactors, the release of tritium and activation products during normal operation as well as after accidents is a significant safety aspect. Calculations have been performed under accidental conditions for unit releases of corrosion products from water coolant loops, of first wall erosion products including different coating materials, and of tritium in its chemical form of tritiated water (HTO). Dose assessments during normal operation have been performed for corrosion products from first wall primary coolant loop and for tritium in both chemical forms (HT/HTO). The two accident consequence assessment (ACA) codes UFOTRI and COSYMA have been applied for the deterministic dose calculations with nearly the same input variables and for several radiological source terms. Furthermore, COSYMA and NORMTRI have been applied for routine release scenarios. The paper analyzes the radioation doses to individuals and the population resulting from the different materials assumed to be released in the environment.D.T.I. Dr. Trippe Ing. GmbH, Karlsruhe.     

A Review of the U.S. Department of Energy's Inertial Fusion Energy Program

This is the final report of a panel set up by the U.S. Department of Energy (DOE) Fusion Energy Sciences Advisory Committee (FESAC) in response to a charge letter from Dr. Ray Orbach (Appendix A). In that letter, Dr. Orbach asked FESAC for an assessment of the present status of inertial fusion energy (IFE) research carried out in contributing programs. These programs include the heavy ion (HI) beam, the high average power laser (HAPL), and Z-Pinch drivers and associated technologies, including fast ignition (FI). This report, presented to FESAC on March 29, 2004, and subsequently approved by them (Appendix B), presents FESAC's response to that charge.     

Source Term and Mobilization Assessment in NSSR-1 of ITER

We describe the radioactive sources in the International Thermonuclear Experimental Reactor (ITER). The most important sources are co-deposited tritium, tritiated water, tokamak dust, and corrosion products. The co-deposited tritium is limited to 1 kg-T; the total on-site tritium inventory in the Basic Performance Phase (BPP) is 4 kg-T. Tritiated water concentrations are kept below 0.2 g-T/m³ in the divertor; other coolant loops have lower tritium concentrations. The in-vessel dust inventory is up to 100 kg-W, 100 kg-Be, and 200 kg-C. The activated corrosion product inventory is kept below 10 kg per loop.     

可加工SiO2气凝胶及其惯性约束聚变靶微柱制备

以正硅酸乙酯（TEOS）为前驱体，采用酸碱两步催化法制备SiO₂醇凝胶。醇凝胶分别经TEOS母液、六甲基二硅胺烷（HMDSA）处理后，采用CO₂超临界干燥法制备出密度在30～100mg/cm³的SiO₂气凝胶。用傅立叶变换红外光谱（FTIR）对疏水性SiO₂气凝胶进行了表征，并用扫描电镜图研究了气凝胶改性前后的微观网络结构。改性后的气凝胶微观骨架变大，部分细小的网络结构消失。改性后的气凝胶在潮湿环境中具有极好的尺寸稳定性和疏水性能。用精密车床加工出了满足惯性约束聚变物理试验要求的ICF靶微柱。     

Study of Fuel Ratios on the Fusion Reactivity in an Inertial Electrostatic Confinement Device Using a Residual Gas Analyzer

Gridded Inertial Electrostatic confinement (IEC) devices are of interest due to their flexibility in burning advanced fuels, their tuning ability of the applied voltage to the reaction cross-section. Although this device is not suitable for power production in its present form, it does have several near term applications. The number of applications of this device increases with increasing fusion reactivity. These devices are simple to operate but are inherently complicated to understand and an effort to incrementally understand the device to improve its operational efficiency is underway at University of Wisconsin, Madison. Of all the parameters under study we are focusing on the effects of flow rate and flow ratio on the fusion reactivity in the present paper. Experiments were conducted to understand the influence of fuel flow ratio on the fusion reactions. The residual gas analyzer (RGA) was used to study the impurity concentration as the flow ratio was changed. It was observed that the higher flow rate resulted in reduced impurity levels and hence an increase in fusion rate. Several different species of gases were detected, some of these molecules formed inside the RGA analyzer. The flow ratio scan revealed that the optimum mixture of D₂ with ³He to be D₂:³He::1:2 for maximum D–³He fusion rate.     

Overview of tritium safety technology at the Tritium Process Laboratory of JAERI

The Tritium Process Laboratory of the Japan Atomic Energy Research Institute is the only laboratory in Japan where grams of tritium can be handled to carry out R&D on tritium processing and tritium safety handling technologies for fusion reactors. The tritium inventory is approximately 13 grams. Since 1988, basic research has been performed using gram-level tritium quantities. During the past 5 years, approximately 1 kilogram of tritium has been handled in experimental apparatus. The total amount of tritium released through the stack of TPL was controlled to less than 1 Ci without any accidents. In order to establish more complete tritium safety for DT fusion reactors, main R&D areas on tritium safety technology at TPL were focused on a new compact tritium confinement system, reliable tritium accounting and inventory control, new tritium waste treatments, and tritium release behavior into a room.     

Fusion and Energy Policy

Presentations from a Fusion Power Associates symposium, Fusion and Energy Policy, are summarized. The topics include an overview of U.S. Department of Energy policies, fusion strategies in Europe and Japan, plans for U.S. participation in the construction of ITER, status of construction of the National Ignition Facility and recent progress in all aspects of magnetic and inertial fusion.