球环型产氚聚变堆概念设计研究

该研究基于球环类型的先进氚生产堆概念而设计,是聚变能发展的中间应用。与传统托卡马克氚生产堆不同,该设计利用球形环的先进等离子体物理性能和紧凑的结构特征,尽量利用真空室内的空间安置氚生产包层以减少氚泄露而增加氚增殖率,相应的堆利用因子为 40%。在二维中子学计算的基础上提出了较为完整的初步概念设计。在逐项分析的基础上对设计的风险、不确定性和后备方案也做了概括的解释。为下一步更详细具体的概念设计提供了直接的依据,具有重要的参考价值。     

聚变实验增殖堆氚系统设计研究(英文)

介绍了核工业西南物理研究院聚变实验增殖堆工程概要设计(FEB-E)中的氚系统设计研究。第一部分介绍包层氚增殖区的划分、几何尺寸、装料特征和用蒙特卡罗程序计算得到的液态锂中的氚浓度分布;第二部分描述根据聚变堆氚物理基础构造的氚循环系统,共分成 10 个子系统及它们之间氚的流程图。运用研制的程序SWITRIM 计算了各个子系统中的氚投料量随时间的变化,满功率运行一年后各个子系统中的氚投料量。研究结果表明起动 143 MW 聚变功率 FEB-E 堆所需要的初始氚投料量大约为 319 g。第三部分对不同的运行状态下的氚泄漏问题进行了分析。潜在的氚泄漏危险可能来自于偏滤器系统从等离子体中抽出的气体。得到的结论是提高FEB-E 堆芯等离子体的燃耗份额从而减少氚的通过量对降低氚的泄漏危险是重要的。     

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。     

聚变堆氚的环境安全评估

对国家863项目聚变实验增殖堆工程概要设计(FEB-E)进行了氚环境安全问题评估。FEB-E是采用液态锂作为包层氚增殖剂，每个包层模块各区之间用隔板隔开．中间通高压氦气冷却、包层第一壁和偏滤器也用氦气冷却。运用自行研制的SWITRIM程序和Sieverts’定律研究了正常工作状态下和事故状态下可能造成氚的环境污染水平。研究表明．正常工作状态下包层液态锂中的氚分压在10^-6～10^-8pa。造成氚环境污染的主要危险来自氚循环回路中的偏滤器子系统的抽出气体泄漏。因此，提高堆芯等离子体燃耗和真空系统设计性能是重要的。     

聚变堆氚材料衡算测量系统研究初探

氘(D)-氚(T)聚变是目前聚变研究的主要对象,也是未来最可能首先实现工程应用的聚变反应.氘氚聚变反应堆在消耗大量氚的同时,需要依靠锂-6增殖足够的氚来实现氚燃料自持.在国内,氚和浓缩锂-6都属于核材料,按照国家核材料管制条例要求,使用和生产核材料须建立核材料衡算系统.为此,本文针对液态包层聚变堆方案,简要归纳了聚变堆...     

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。     

聚变反应堆含氚废水处理系统的设计

以联合电解催化交换-气相色谱(CECE-GC)为技术路线基础,对聚变反应堆(International thermonuclearexperimental reactor,ITER)含氚废水处理系统(Water detritiation system,WDS)进行了总体设计和主要子系统的设计.与目前的重水提氚演示系统相比,ITER-WDS的不同之处在于不使用氢氧复合器,不采用碱性电解池而使用固体聚合物电解池(Solid polymer electrode,SPE),增加了Pd/Ag膜渗透系统进行氚的回收.     

基于积分分析方法对聚变堆启动氚量与所需氚增殖比的评估

托卡马克聚变堆的主要发展方式包括混合堆、纯聚变堆。关于托卡马克聚变堆氚自持的研究,国内外主要采用平均滞留时间方法进行研究,并且针对聚变功率较低的混合堆的氚自持研究较少。本工作采用更符合实际的积分分析方法分析了混合堆、纯聚变堆氚自持的启动氚量、氚增殖比(TBR)要求。研究结果表明：启动氚量、备用氚量与聚变功率具有线性关系,所需TBR与聚变功率呈反比例关系;混合堆聚变功率较低,所需TBR较高,工程实现所需TBR挑战较大,需要通过限制长期氚滞留量以降低所需TBR要求;纯聚变堆聚变功率高,所需TBR较低,工程实现所需TBR挑战较小,但备用氚需求达数十千克,应考虑氚系统的冗余设计或提高氚系统的可靠性、可维护性,以降低备用氚的使用规模;运行因子是聚变堆的一个重要设计指标,在此着重分析了运行因子对所需TBR的影响,并重新定义了一个聚变堆氚自持的关系式,以突出运行因子对氚自持的重要影响。     

聚变-裂变混合堆设计研究

利用MCNP5和MONK9A程序对聚变驱动裂变混合堆进行了初步研究,在等离子体第1壁外侧依次包覆长方体形状的燃料组件和产氚组件,形成裂变堆芯包层和产氚区.对分别装载贫铀、天然铀、贫铀MOX和天然铀MOX等4种燃料的混合堆进行了研究分析,其中,后两种燃料在整个运行寿期内的功率放大倍数和氚增殖比满足设计要求.通过随燃耗变化的同位素含量分析,初步探讨了混合堆的铀-钚燃耗循环策略.     

氚化学与氚分析进展与展望

氚是聚变堆的重要燃料,氚的问题是制约聚变能源发展的瓶颈问题之一。氚化学中的科学技术问题是解决涉氚工艺的基础,氚分析测量是氚操作中获取氚信息的主要途径。本文综述了近年来氚化学与氚分析方面的研究进展,从氚与材料相互作用、涉氚材料中的氦行为、氚的氢同位素效应、氚的辐射生物效应,以及氚的分析测量方法五个方面对研究进展进行了分析,并对聚变堆中所面临的挑战进行了展望。     

空气中氡甄别型氚监测仪的研制

介绍了一种可甄别空气中惰性气体氡的高灵敏度氚实时连续监测仪,其一体化的平行板型丝壁测氚电离室和镀铝聚酯膜窗平行板型测氡密闭电离室,实现了空气中氚监测过程中氡干扰的实时、同步测量与补偿。该氡甄别型氚监测仪不仅在一般使用环境条件下减少了环境氡的影响．提高了氚监测仪的探测灵敏度,而且可在氡浓度比环境高100倍的场所探测到100Bq／L的氚浓度。对于监测环境中其他8放射性惰性气体．这种新型的氚监测仪预期将会有更好的甄别补偿性能。     

Economic analysis of a magnetic fusion production reactor

The magnetic fusion reactor for the production of nuclear weapon materials, based on a tandem mirror design, is estimated to have a capital cost of $1.5 billion and to produce 10 kg of tritium/year for $22,000/g or 940 kg/year of plutonium in the plutonium mode for $250/g plus heavy metal processing. A tokamak-based design is estimated to cost $1.5 billion and to produce 10 kg of tritium/year for $29 thousand/g. For comparison, a commercially sized tandern mirror fusion breeder selling excess electricity and fissile material to commercial markets is estimated to cost $3.6 billion and to produce tritium for $2.6 thousand/g and plutonium for $34/g plus heavy metal processing.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.     

Nuclear design and analysis of a magnetic fusion production reactor

Tandem-mirror- and tokamak-based magnetic fusion production reactors are predicted to have tritium breeding ratios of 1.67 and 1.49, respectively. The latter value replaces one (1.56) that is used elsewhere in the sequence of papers in this issue. Blanket energy multiplication for both is predicted to be about 1.3. With the tandem mirror operating in the plutonium production mode, the net plutonium-plus-tritiurn breeding ratio is 1.74. Blanket energy multiplication for the plutonium mode is predicted to be 2.4 at a plutonium-uranium ratio of 0.7% and a uranium volume fraction of 3%.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.     

Enhanced tritium production for fusion reactors via3He(n,p)3 H in the heavy water moderator of a CANDU reactor

There is an increasing requirement for tritium to supply the fuel needs of current experimental fusion devices and in the initial startup of future power generating reactors. Tritium is produced in heavy water reactors through deuterium activation, but the total production capacity of Canadian operated CANDUs will fall short of future demands, during the period before and for some time after self-sufficient reactors become available. Consequently, methods of enhancing tritium generating rates warrant investigation. Herein we provide the results of an inquiry into the feasibility of enhancing tritium production levels through the activation of helium-3 following its external addition to the heavy water moderator system of a hypothetical 500–600 MWe CANDU reactor. The approach adopted involves simulation of the temporal evolution of the tritium activities, originating from²H(n,)³H and³He(n, p)³H, as described by a simple first order kinetic model. The results suggest that the frequent addition of helium-3 to the moderator water will enhance tritium production inventories. The enhancement factor is highly dependent upon the rate at which helium-3 irretrievably escapes to the moderator cover gas. However, the direct activation of helium-3, contained in a closed loop such as the annulus gas system, for example, would be essentially complete within a few weeks without any significant loss.     

Neutron emission and the tritium content associated with deuterium-loaded palladium and titanium metals

An experimental investigation has been conducted on samples of palladium and titanium metals which have been loaded with deuterium through the electrolysis of D₂O and by absorption of D₂ gas. In approximately 200 experiments on 25 cells, statistically significant evidence for neutron emission was obtained in three separate experiments from one palladium cathode. Observed rates are 3–4 times the background rate and correspond to source strengths up to 50 neutrons/min. The pulse height response of the NE213 liquid scintillator-based detectors corresponds to that expected for 2.45 MeV neutrons. Tritium has been identified in nine Pd−Ni electrolytic cells, at levels corresponding 10¹²–10¹⁶ atoms. Activity buildup curves indicate that the apparent production occurs over a time period as short as a few hours. This work supported by the Office of High Energy and Nuclear Physics of the U.S. Department of Energy under Grants DE-FG05-86ER40256 and DE-FG05-88ER40437     

Selection of a toroidal fusion reactor concept for a magnetic fusion production reactor

The basic fusion driver requirements of a toroidal materials production reactor are considered. The tokamak, stellarator, bumpy torus, and reversed-field pinch are compared with regard to their demonstrated performance, probable near-term development, and potential advantages and disadvantages if used as reactors for materials production. Of the candidate fusion drivers, the tokamak is determined to be the most viable for a near-term production reactor. Four tokamak reactor concepts (TORFA/FED-R, AFTR/ZEPHYR, Riggatron, and Superconducting Coil) of approximately 500-MW fusion power are compared with regard to their demands on plasma performance, required fusion technology development, and blanket configuration characteristics. Because of its relatively moderate requirements on fusion plasma physics and technology development, as well as its superior configuration of production blankets, the TORFA/FED-R type of reactor operating with a fusion power gain of about 3 is found to be the most suitable tokamak candidate for implementation as a near-term production reactor.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 from its 1982 version (except for Tables II and III and Fig. 1), explaining the fact that some of the material is dated.     

氦中痕量杂质气体气相色谱检测分析方法研究

为实现聚变堆氘氚燃料工艺气中痕量杂质气体组分的快速检测分析,需建立特殊的高精度在线气相色谱检测分析方法。以高纯氦作为载气,在不同的色谱柱温度和载气流速控制下,通过分子筛毛细管柱和PLOT-Q柱进行分离,采用放电氦离子化检测器（DID）进行检测,对氦中含量为1、10以及100 ppm的杂质标准气体进行检测分析。结果表明：在柱温为40 ℃、流速为15～20 mL/min实验条件下,分子筛柱在160 s内能够实现H₂、O₂、N₂、CH₄和CO全部分离,且柱效较高,响应值的重复性较好,H₂和O₂之间的分离度高于1.5,实现了完全分离;在柱温为40 ℃、流速为20 mL/min时,PLOT-Q柱分离CO₂组分效果最佳。