磁聚变开发的球形环途径

低环径比托卡马克或球形环途径可能具有使磁约束聚变从政府拨款的研究规划转变到专业市场所需的两个关键要素：进入市场的装置费用低，功率小和尺寸小，扩大成较魇装置具有很强的经济性。     

SUNIST球形托卡马克的研究进展

球形托卡马克为聚变能的商业应用提供了一条可能的途径。中国联合球形托卡马克SUNIST以真空室的环向和极向都有绝缘隔缝为结构特征。该装置的主要任务是研究低环径比等离子体的基本特性和等离子体的非感应加热与电流驱动。包括同轴磁螺旋性注入电流启动、电极放电辅助电子回旋波电流启动、电子伯恩斯坦波以及离子高次谐波快波加热与电流驱动。装置已经顺利组装完毕,并安装了磁测量、静电探针和软X射线等基本的诊断系统。目前正处于系统联调阶段。     

磁约束聚变装置辐射监测系统概况

采用氘、氚燃料的核聚变反应会产生大量的中子、γ射线及活化产物等,对人员和环境的辐射安全产生影响。为了减小电离辐射带来的影响,需要准确掌握聚变装置核辐射场强度的时间与空间分布信息。世界上已建设的磁约束聚变装置,均根据其自身运行工况特点,建立了完整的核辐射监测系统来应对电离辐射带来的潜在影响。通过对磁约束聚变装置运行及维护期间辐射剂量的监测,获得实验场所与外围环境的电离辐射和放射性核素数据,为辐射安全防护管理提供数据支撑。基于对国内外磁约束聚变装置辐射监测系统的调研,本文归纳了此类装置主要的电离辐射源项及监测系统架构,进而介绍了磁约束聚变中子与γ辐射剂量的测量方法及常用探测器。最后综述了国内外核聚变装置辐射监测系统的研究状况,展望了未来核辐射监测系统的发展趋势与目标。     

在Windows 95下开发多发弹丸发射控制程序

在受控核聚变研究中,氢同位素固态弹丸的高速注入已成为核聚变加料实验的热点。因其具有加料效率高、能形成芯部高度峰化的密度分布、拓宽装置运行区域以及改善等离子体约束性能等优点,因而世界各国在磁约束核聚变装置上广泛采用这种加料技术。为此由中、俄双方联合研制了用于HL-1M装置的多发弹丸加料系统。该系统工作于强电场、强磁场环境下,弹丸注入时刻为毫秒量级控制,从而决定了其发射控制只能采用计算机自动控制方式。     

世界受控聚变研究的现状和趋势

聚变能是轻原子核聚变产生的能量,是一种取之不尽的、清洁的、廉价的和环境安全的能源,是当今人类开发的最先进的能源。 1952年试制成功的氢弹实际上就是人类开发的第一个核聚变装置,但不是受控装置,只能用于军事。它是惯性约束的内爆形式,利用核裂变去引发核聚变释放聚变能。而人类的最终目标是实现受控聚变。 核聚变的研究途径多种多样,但由于种种原因,目前主要集中在磁约束聚变和惯性     

HL-1M等离子体加料密度特性

等离子体加料和密度控制是磁约束核聚变基本研究内容之一。HL-1M实验装置用8发PI系统与SMBI和GP组成联合加料系统,以它们相互配合进行了一系列放电实验,取得了丰硕的成果,本文就等离子体电子密度、改善约束特性与燃料粒子注入深度、放电装置器壁再循环的关系等结果作一介绍。     

我国磁约束核聚变标准化工作探讨

介绍了在当前磁约束核聚研究中使用现有技术标准,开展标准化工作带来的优点.以国际热核聚变实验堆(ITER)计划为例,分析了ITER计划中技术标准的总体应用情况、特定部件的标准化情况,以及我国已签约ITER采购包采购安排协议(PA)中引用标准的特点.概述了我国磁约束核聚变标准化工作现状,并给出了当前我国磁约束核聚变标准化工...     

磁约束核聚变能源开发的进展和展望

开发利用核聚变能源是人类面临的最具挑战性的世纪难题 ,人类为此奋斗了半个世纪。在过去的 5 0年中 ,我们取得了极大进展。今天 ,我们实际上已经踏进建造聚变能源装置的门坎。磁约束核聚变研究以托卡马克装置为主流 ,在追求等离子体高温、高密度和高约束模式的不懈努力中 ,近年取得突破性进展 ,实现了聚变功率输出和科学意义上的“能量自持”。一个以“点火和自持燃烧”为研究目标的国际热核实验堆 (即ＩＴＥＲ计划 )已经完成工程设计 ,为降低造价和压缩规模而提出的替代方案 (ＲＣ ＩＴＥＲ)也已经出台。惯性约束核聚变研究取得的成就也不…     

可控核聚变与国际热核实验堆（ITER）计划

介绍了我国能源的基本隋况,核聚变能和可控核聚变的基本原理,以及国际热核聚变实验堆ITER的历史与现状。对我国磁约束核聚变的研究发展历程做了简要的回顾。     

美国核聚变研究增加经费加快步伐

【美国《今日物理》1980年5月号报道】为加快核聚变研究步伐,美国能源部将使1981年财政年度的磁约束和惯性约束聚变研究的经费增加9％。磁约束1981年财政年度磁约束聚变研究的财政预算为3.96亿美元,比1980年增加了11％。众议院能源研究和开发小组委员会认为,加快聚变研究步伐的时机已经     

Low Radioactivity Fusion Reactor Based on the Spherical Tokamak with a Strong Magnetic Field

The most important feature of the spherical tokamak is the possibility of high-β plasma confinement (β is the ratio of plasma pressure to magnetic field pressure). So, spherical tokamak can be considered as a possible confinement system for D–³He fusion reactor. Present paper study the ability to develop powerful D–³He reactor based on a spherical tokamak (fusion power about 3 GW). The following parameters are considered as optimization criteria: (1) the ratio of confinement time to the value predicted by ITER98y2 scaling; (2) the neutron flux from the plasma.     

Corrected Formulation for Estimation of Ripple in Large Aspect Ratio Tokamaks

The finite number of toroidal field coils of a tokamak destroys the perfect axisymmetry of the device. The coils produce a short wavelength ripple in the toroidal magnetic field strength as a field line follows round the torus, which becomes important in transport and confinement properties of plasma. Hence, a quick and accurate estimation of ripple becomes important. We have noticed that a previously reported analytical formulation by Princeton Plasma Physics Laboratory team is not applicable to our large aspect ratio tokamak, and have devised a slightly modified form which has greatly improved the accuracy of the analytical fit.     

Three Game Changing Discoveries: A Simpler Fusion Concept?

This paper encourages exploration of a broad range of magnetic fusion concepts in parallel with mainline tokamak development. Such exploration will certainly lead to increased understanding of fusion science and possibly to an attractive fusion energy concept. As an example, this paper describes three discoveries which greatly increase the attractiveness of the magnetic mirror plasma confinement concept. The mirror concept is thought to have three unattractive characteristics. The magnets are complex, the plasma is plagued with micro-instabilities and the electron temperature would never approach required keV levels. Persistent research on the gas dynamic trap device at the Budker Institute of Nuclear Physics in Russia and elsewhere have overcome these three deficiencies. Stable high energy density plasma can be confined with simple circular magnets, micro-instabilities can be tamed, and electron temperatures reaching a keV have been measured. These three accomplishments provide a basis to reconsider the mirror concept as a neutron source for medical applications, fusion materials development, nuclear fuel production, and fusion energy production.     

The structural design of superconducting magnets for the large coil program

Fusion reactor designs based on magnetic confinement will require the use of superconducting magnets to make them economically viable. For a tokamak fusion reactor; large magnetic field coils are required to produce a toroidal magnetic confinement volume. Although superconductors have been used for approximately 20 years, several requirements for their application in fusion reactors are beyond demonstrated technology in existing magnets. The Large Coil Program (LCP) is a research, development, and demonstration effort specifically for the advancement of the technologies involved in the production of large superconducting magnets. This paper presents a review of the status of the structural designs, analysis methods, and verification tests being performed by the participating LCP design teams in the US, Switzerland, Japan, and the Federal Republic of Germany. The significant structural mechanics concerns that are being investigated with the LCP are presented.     

A Consideration on Increasing Current Density in Normal Conducting Toroidal Field Coil for Spherical Tokamak Power Plant

The center post is the most critical component as an inboard part of the toroidal field coil for the low aspect ratio tokamak. During the discharge it endures not only a tremendous ohmic heating owing to its carrying a rather high current but also a large nuclear heating and irradiation owing to the plasma operation. All the severe operating conditions, including the structure stress intensity and the stability of the structure, largely limit the maximum allowable current density. But in order to contain a very high dense plasma, it is hoped that the fusion power plant system can operate with a much high maximum magnetic field BT ≥12 T-15 T in the center post. A new method is presented in this paper to improve the maximum magnetic field up to 17 T and to investigate the possibility of the normal conducting center post to be used in the future fusion tokamak power plant.     

Hybrid Fusion: The Only Viable Development Path for Tokamaks?

The world needs a great deal of carbon free energy, and soon, for civilization to continue. Fusion’s goal is to develop such a carbon free energy source. For the last 4 decades, tokamaks have been the best magnetic fusion has to offer. But what if its development stops short of commercial fusion? This paper introduces ‘conservative design principles’ for tokamaks. These are very simple, are reasonably based in theory, and have always constrained tokamak operation. Assuming they continue to do so, it is unlikely that tokamaks will ever make it as commercial reactors. This is independent of their confinement properties. However because of the large additional gain in hybrid fusion, tokamaks reactors look like they can make it as hybrid fuel producers, and provide large scale power by mid century or shortly thereafter.     

Magnetic fusion commercial power plants

Toroidal magnetic systems offer the best opportunity to make a commercial fusion power plant. They have, between them, all the features needed; however, no one system yet meets the ideal requirements. The tokamak is the most advanced system, and the proposed International Thermonuclear Experimental Reactor (ITER) and Tokamak Physics Experiment (TPX) will build upon the existing program to prepare for an advanced tokamak demonstration plant. Complementary toroidal systems such as the spherical torus, stellarator, reversed-field pinch, field-reversed configuration, and spheromak offer, between them, potential advantages in each area and should be studied in a balanced fusion development program.     

Smaller & Sooner: Exploiting High Magnetic Fields from New Superconductors for a More Attractive Fusion Energy Development Path

The current fusion energy development path, based on large volume moderate magnetic B field devices is proving to be slow and expensive. A modest development effort in exploiting new superconductor magnet technology development, and accompanying plasma physics research at high-B, could open up a viable and attractive path for fusion energy development. This path would feature smaller volume, fusion capable devices that could be built more quickly than low-to-moderate field designs based on conventional superconductors. Fusion’s worldwide development could be accelerated by using several small, flexible devices rather than relying solely on a single, very large device. These would be used to obtain the acknowledged science and technology knowledge necessary for fusion energy beyond achievement of high gain. Such a scenario would also permit the testing of multiple confinement configurations while distributing technical and scientific risk among smaller devices. Higher field and small size also allows operation away from well-known operational limits for plasma pressure, density and current. The advantages of this path have been long recognized—earlier US plans for burning plasma experiments (compact ignition tokamak, burning plasma experiment, fusion ignition research experiment) featured compact high-field designs, but these were necessarily pulsed due to the use of copper coils. Underpinning this new approach is the recent industrial maturity of high-temperature, high-field superconductor tapes that would offer a truly “game changing” opportunity for magnetic fusion when developed into large-scale coils. The superconductor tape form and higher operating temperatures also open up the possibility of demountable superconducting magnets in a fusion system, providing a modularity that vastly improves simplicity in the construction, maintenance, and upgrade of the coils and the internal nuclear engineering components required for fusion’s development. Our conclusion is that while tradeoffs exist in design choices, for example coil, cost and stress limits versus size, the potential physics and technology advantages of high-field superconductors are attractive and they should be vigorously pursued for magnetic fusion’s development.     

Influence of the safety factor profile on the particle and heat transport in the Globus-M spherical tokamak

The advanced tokamak scenario is a promising operation scenario for ITER and fusion neutron sources. In this scenario the minimum value of the safety factor in the center of the plasma exceeds unity. In the compact spherical tokamak Globus-M, the formation of such conditions is possible with neutral beam injection at the current ramp-up phase. Due to the slower diffusion of current inside the plasma, a zone is formed with reduced heat and particle transport across the magnetic field, which affects the temperature and density profiles of the plasma. This leads to the peaked density profile formation and improvement of the energy confinement time. To achieve a high fraction of the bootstrap current, it is necessary to increase the plasma pressure. At the same time, the maximum allowable pressure is limited to the normalized beta limit.