负离子——中性束注入系统发展概况

本文综述了大型和下一代受控核聚变装置对负离了－中性束系统的技术以及负离子－中性束注入器的发展概况２，阐明了所取得的进展和需要进一步解决的物理和技术问题。     

用于ITER中性束注入器的大体积负离子源的改进

对负离子源进行了改进以实现用于国际热核实验堆（ITER）的大功率中性束注入器，已在植入铯的多会切等离子体发生器中在0．1Pa的非常低的气压下产生了31mA.cm^2-(H^-)的强负离子电流密度，该等离子体发生器与ITER源有着相同的概念，对于一个真空绝缘的加速器，已完成了达1．8米的长距离真空间隙的耐压实验，已表明从真空击穿到气体放电的气压距离乘积（pd)的过渡区约为0．2Pa.m，这离ITER源的运行区足够高了，在试验和实验的基础上建造了真空绝缘加速器样机，成功地证实了高能H^-束加速达到970keV,37mA,1s。     

中性束注入器水流热量计系统优化分析

介绍了用于计算中性束注入实验中束流功率沉积的水流热量计系统优化及优化后的测试结果。前期的水流热量计系统利用串口传输和分散式牛顿模块采集,采样率低、传输速度慢、抗干扰能力差、无法满足实验要求;优化后的系统基于虚拟仪器,采用TCP\IP协议传输和虚拟仪器技术,提高了系统的采样率和精度、优化了数据传输速度及抗干扰能力。优化后的系统经测试可以准确监测中性束注入器装置上各热承载部件冷却水的温升及流量,并分析得到中性束注入时束流在各热承载部件上的功率沉积。实验结果表明优化后的系统工作稳定,使用灵活,数据准确,满足实验要求。     

EAST中性束注入器稳态偏转磁铁的参数设计与估算研究

中性束注入器偏转磁铁是剥离束流中剩余离子的关键设备,它与剩余离子吞食器等内部部件构成了中性束注入器的束偏转系统。束偏转系统的性能对中性束注入器束流的品质及其束传输效率发挥着重要作用。本文根据EAST(Experimental Advanced Superconducting Tokamak,EAST)中性束注入器对束偏转系统的要求,对其偏转磁铁各性能参数进行了估算。为中性束注入器设计了一台用以剩余离子180°偏转的偏转磁铁。该偏转磁铁采用H型二极电磁铁结构;其磁极端面设计为138cm×47cm的圆角矩形结构;其线圈设计为每侧2饼,每饼2层,每层8根的串联结构,导线选用外方内圆空心铜导体,以满足偏转磁铁稳态运行的需要。该设计的偏转磁铁在370 A励磁电流条件下,可提供80keV氘离子束偏转所需的磁场。实验测试结果显示:500 A励磁电流稳态运行条件下,偏转磁铁线圈冷却水温升约21.5℃,该设计结构的偏转磁铁满足EAST中性束注入器满参数稳态运行和未来运行参数逐步提高的需要。     

在用于大型螺旋装置中性束注入系统的1／3比例氢负离子源中Cs沉积的优化

利用一个紧凑的铯沉积系统对用于大型螺旋装置-中性束注入（LHD-NBI）系统的1/3比例氢负离子源的内表面进行直接的沉积,试验了在3-200mg范围内小的、很确定的铯量的沉积,在纯氢运行模式和有铯模式下都进行了负离子的引出和加速。对等离子体室单纯的3-30mg的铯沉积使H∧-产额暂时增高2-5倍,但是在几个放电脉冲之内该产额又降低到原先的稳态值。在3-5h/60次放电间隔之内连续两次的30mg沉积,产生了类似的H∧-的瞬时增长,但达到很大的H∧-产额的稳态值。在20-120h/150-270次放电的间隔内,更大量的0.1-0.1gGs的沉积,在一个长的运行周期（2-5d）内改善了H∧-产额。对等离子体室各个壁的定向的Cs沉积表明了近似同样的H∧-的增加。对被水泄漏污染的表面沉积0.13gGs产生了与单一30mg铯沉积时相类似的H∧-的瞬时增加和H∧-的稳态水平,用铯等离子体的0.1g沉积得到了用相同量的铯原子沉积时获得的H∧-产额的一半。与在相同的放电功率下12根灯丝运行相比,在8根灯丝的放电运行期间记录到更高的稳态H∧-流值和更小的H∧-产生速率。     

ASDEX中性束注入器的供电系统

本文介绍ＡＳＤＥＸ中性束注入器供电系统概况，对注入器的高压电源，磁场反馈控制和弧打坑电路作较较详细的介绍。     

HL-1M 中性束注入器快速断电保护器

介绍了用于ＨＬ－１Ｍ中性束注入器电源系统中的一个快速断电保护器。它能快速探测注入器过黉,过流和打火击穿等运行故障并给出故障信号,以便快速分断电源,保护离子源和电源本身免于损坏。详细介绍了电路结构,工作原理和实验结果。保护器在强线路干扰和电磁干扰下可靠工作。     

A preliminary conceptual design study for Korean fusion DEMO reactor

As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy, so called fast-track approach. Korean strategy for fusion energy can be regarded as a fast-track approach and one special concept discussed in this paper is a two-stage development plan. At first, a steady-state Korean DEMO Reactor (K-DEMO) is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used as a component test facility. Then, at its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electric generation on the order of 300 MWe and the competitiveness in cost of electricity (COE). The major radius is designed to be just below 6.5 m, considering practical engineering feasibilities. By using high performance Nb₃Sn-based superconducting cable currently available, high magnetic field at the plasma center above 8 T can be achieved. A design concept for TF magnets and radial builds for the K-DEMO considering a vertical maintenance scheme, are presented together with preliminary design parameters.     

A tritium permeation model for conceptual fusion reactor designs

A transient tritium permeation model is developed based on a simplified conceptual DT-fueled fusion reactor design. The major design features described in the model are a solid breeder blanket, a low pressure purge gas in the blanket, and a high pressure helium primary coolant. Tritium inventory in the breeder is considered to be due to diffusive hold-up and solubility effects. It is assumed that diffusive hold-up is the dominant factor in order to separate the solution for the breeder tritium concentration. The model was applied to the STARFIRE-Interim Reference Design, whose system parameters yielded a breeder tritium inventory on the order of grams, based on an average pellet radius of 10^?3 cm. The breeder pellets reach their steady-state tritium content in approximately 1.4×10⁴ s from system start-up, assuming continuous full power operation. Both the steady-state breeder tritium concentration and the time to reach that steady-state are proportional to the pellet radius squared. Other candidate solid breeders were considered, and their effect on the blanket tritium inventory was noted. The addition of oxygen to the primary coolant loop was required in order to keep the tritium losses through the heat exchanger to within the design goal of 0.1 Ci/day.     

A conceptual design of light ion beam fusion reactor—UTLIF(1)

UTLIF(1) is a conceptual design study on a light ion beam driven fusion reactor based on a concept of pin-bundle blanket. The study includes nuclear and structural analyses of the blanket, consideration on materials, tritium handling system and power conversion system designs, pellet and beam driver designs, and economic analysis of the plant. The pin-bundle blanket has been shown to be attractive for light ion beam fusion reactors. Some subjects to be developed have been pointed out from reactor engineering aspects.     

新型反应堆可视化设计方法研究

以核聚变反应堆为研究对象,提出了一种参数化的零部件设计为先导,以可变异的零部件组装逻辑为手段的新型的可视设计方法,所有的设计对象采用实体造型技术,可对系统的设计结果进行三维渲染和动态仿真。设计的对象包括偏滤器、中心柱、包层、纵向场、极向场等,以偏滤器为例,部件组成又包括轨道、支体、外靶板、内靶板、能量沉积板、圆顶和翼。所介绍的设计系统为核聚变装置的灵活设计提供了一种有效的手段,同时也将促进核聚变装     

The UWMAK-II study: A conceptual design of a helium cooled, solid breeder, Tokamak fusion reactor system

UWMAK-II is a conceptual design study of a low ß, circular Tokamak fusion power reactor. The aim of the study has been to perform a self-consistent analysis of a probable future fusion power system based on the philosophy that design decisions, wherever possible, should be conservative and should be based on present technology. As such, this system will not be the smallest, the least expensive, or the optimum Tokamak reactor. Rather, it represents a feasible system which we use to assess the technological problems uncovered and to examine possible solutions. The plasma is designed to generate 5000 MW(th) during a pulse and 1709 MW(e) continuously based upon a burn cycle with a 90 min burn and a 6.5 min rejuvenation period. The plasma carries a current of 14.9 MA and is designed with a double null poloidal divertor for impurity control and particle pumping. In addition, a low Z liner in the form of a carbon curtain is included to eliminate any source of high Z impurities. Plasma heating to ignition involves the use of neutral beam heating for a 10 sec period during which 200 MW of 500 keV deuterium atoms are injected into the plasma.The blanket design employs helium cooling and the solid lithium-bearing compound, lithium aluminate (Li₂Al₂O₄) for breeding tritium. The structural material is 316 stainless steel and beryllium is used as a neutron multiplier. The neutron radiation environment produces radiation damage that considerably influences blanket and system performance. The most significant effect is the loss of ductility which appears to limit the usable lifetime of the blanket first wall to about 2 yr at a 14 MeV neutron wall loading of 1.16 MW/m². The solid breeder blanket minimizes the tritium inventory but because of the low fractional burnup in the plasma and the need for roughly a one day reserve of fuel, the inventory is 17.7 kg. Induced radioactivity levels in the structure are of the order of 1 Ci/W(th) at shutdown after two years of operation. The main contributors to the activity are ) and ). Afterheat levels are slightly above 1% of thermal power but the afterheat power density is low, less than 0.1 w/g. The power cycle involves a He---Na intermediate heat exchanger followed by a sodium—steam system. The sodium intermediary is used to minimize tritium leakage through the power cycle and to provide a working fluid for thermal energy storage such that continuous electrical output is produced despite a pulse plasma cycle. A materials resource study has been completed for a UWMAK-II type system and beryllium appears to present a particular problem with regard to adequate resources. Other materials that could present problems of procurement include chromium and nickel. A preliminary economic analysis has been carried out to identify major cost areas and this is described.     

Nuclear design for a heavy ion-beam fusion reactor,HIBLIC

A nuclear analysis was carried out for a heavy ion-beam fusion reactor, HIBLIC. The analysis includes the target and chamber neutronics, time-dependent radiation damage in the first wall, and radiation streaming through beam ports. It is found that the reactor chamber is characterized by its high tritium breeding ratio, low radiation damage in the second wall, and low induced activity. To reduce the radiation damage in the superconducting, focusing magnets, tapering the beam ports along the direct line-of-sight component of the source neutron is necessary in the magnet regions and also in the collimator region.     

Nuclear design of a heliotron-H fusion power reactor

Nuclear analysis was carried out for the heliotron-H fusion power reactor employing anl=2 helical heliotron field. The neutronics aspects examined were (a) tritium breeding capability, (b) shielding effectiveness for the superconducting magnet (SCM), and (c) induced activity after shutdown. In this reactor design of the heliotron-H, the space available for the blanket and shield is limited due to the reactor geometry. Thus, some parametric survey calculations were performed to satisfy the design requirements. The nucleonic design features of the heliotron-H are as follows. An adequate tritium breeding ratio of 1.17 is obtained when a 10-cm thick Pb neutron multiplier and a 40-cm thick Li₂O breeding blanket are used. In this case, the total nuclear energy deposition is 16.10 MeV per 14.06 MeV incident neutron. The performance of the SCM is assured during 2 yr of continuous operation using a 20-cm thick tungsten shield. Biological dose rate behind the SCM at 1 day after shutdown is too high for hands-on maintenance.