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。     

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。     

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。     

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。     

聚变堆氚分析程序TAS1.0的设计与开发

氚是聚变堆的重要燃料之一,对聚变堆氚系统进行分析从而实行有效的氚控制是聚变研究的重要内容之一.在中国系列液态金属锂铅包层聚变堆概念设计研究基础上,利用现代软件工程方法及面向对象技术设计思想,发展了聚变堆氚分析程序TAS1.0,可用于聚变堆氚自持分析、氚燃料管理及氚安全性分析与研究,并可为聚变堆包层及燃料循环系统设计与分析提供技术支持.通过一系列的测试校验,表明了该程序的正确性与有效性.本文主要介绍该程序的系统设计、技术特点与程序测试.     

聚变堆失真空事故下氚的大气扩散模拟研究

聚变反应堆发生失真空事故的情况下氚会泄漏到环境中,氚大气扩散模拟是聚变堆事故后果评价的重要内容。基于高斯烟团模型以及Pasquill稳定度分类方法,考虑重力沉降、烟气抬升、风速等因素的影响,建立了适用于事故下瞬态分析的大气扩散模型,在高斯烟团模型中加入修正了像源贡献的地面反射系数,提高了模型对于地面边界处干沉降的计算效果。选取加拿大氚气释放实验和美国萨凡那河工厂氚释放事故验证了所建立模型的准确性,模型的计算结果与HotSpot 3.0和UFOTRI软件的精度相当。选取国际热核聚变实验堆（International Thermonuclear Experimental Reactor,ITER）的失真空事故作为研究对象,分析了氚的分阶段释放、风速以及释放高度对氚扩散分布的影响。结果表明：氚的分阶段释放会导致沿下风向出现两个高放射性区域;释放高度和风速的增加会强化氚在大气中的扩散行为,从而减弱放射性在近场的积聚。     

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...     

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.     

美国对重新起动产氚堆的时间表有争议

【美国《华盛顿邮报》1991年2月25日报道】据总审计局(GAO)的一份研究报告说,美国能源部有足够的氚来满足今后若干年的核武器生产需要,可大大放慢萨凡纳河     

聚变堆安全特性评价研究

确保核安全是未来聚变堆设计、建造和运行过程中必须坚持的最高原则,是聚变堆获得建造和运行许可的前提条件,也是聚变能得以吸引公众的主要理由之一。聚变堆具有高能中子、大量放射性氚、复杂结构、极端服役环境等特点,具有独特的潜在安全问题,因而必须开展针对性研究。本文将从聚变中子与放射性源项、热流体与能量传输、氚安全与环境影响、可靠性与风险管理、安全理念与公众接受度五个方面分别总结其安全特性,系统梳理其关键技术挑战,为建立聚变安全评价体系提供技术支持,进而服务于未来聚变堆的设计与建造。     

Modification of PROMETHEUS Reactor as a Fusion Breeder and Fission Product Transmuter

This study presents the analyses of the fissile breeding and long-lived fission product (LLFP) transmutation potentials of PROMETHEUS reactor. For this purpose, a fissile breeding zone (FBZ) fueled with the ceramic uranium mono-carbide (UC) and a LLFP transmutation zone (TZ) containing the ⁹⁹TC and ¹²⁹I and ¹³⁵Cs isotopes are separately placed into the breeder zone of PROMETHEUS-H design. The neutronic calculations are performed by using two different computer codes, the XSDRNPM/SCALE4.4a neutron transport code and the MCNP4B Monte Carlo code. A range of analyses are examined to determine the effects of the FF, the fraction of ⁶Li in lithium (Li) and the theoretical density (TD) of Li₂O in the tritium breeder zone (TBZ) on the neutronic parameters. It is observed that the numerical results obtained from both codes are consistent with each other. It is carried out that the profiles of fission power density (FPD) are flattened individually for each FF (from 3 to 10%). Only, in the cases of FF ≥ 8%, the system is self sufficient from the point of view of tritium generation. The results bring out that the modified PROMETHEUS fusion reactor has capabilities of effective fissile breeding and LLFP transmutation, as well as the energy generation.     

Effect of Lithium Enrichment on the Tritium Breeding Characteristics of Various Breeders in a Fusion Driven Hybrid Reactor

Selection of lithium containing materials is very important in the design of a deuterium–tritium (DT) fusion driven hybrid reactor in order to supply its tritium self-sufficiency. Tritium, an artificial isotope of hydrogen, can be produced in the blanket by using the neutron capture reactions of lithium in the coolants and/or blanket materials which consist of lithium. This study presents the effect of lithium-6 enrichment in the coolant of the reactor on the tritium breeding of the hybrid blanket. Various liquid–solid breeder couples were investigated to determine the effective breeders. Numerical results pointed out that the tritium production increased with increasing lithium-6 enrichment for all cases.     

Impact of Solid Breeder Materials on Tritium Breeding in a Hybrid Reactor

Tritium breeding ratio (TBR) is one of the important parameters in design of a Deuterium–Tritium (DT) driven hybrid reactor. Therefore, selection of tritium breeder materials to be used in the blanket is very crucial. In this study, tritium breeding potential of the solid breeders, namely, or in a (DT) fusion driven hybrid reactor fuelled with or was investigated. For this purpose in addition to these solid breeders, different types of liquid breeders, namely natural lithium, Flibe, Flinabe and were used to examine the tritium breeding behavior of liquid–solid breeder couple combinations. Numerical calculations were carried out by using Scale 4.3. According to numerical results, the blanket with fuel using natural lithium as coolant and as solid breeder had the highest TBR value.     

基于PHA的TBM氚系统安全分析

使用预先危险性分析(PHA)方法对国际热核聚变实验堆计划(ITER)实验包层(TBM)附属氚处理系统的初级氚包容系统进行安全分析。首先给出PHA分析的基本过程,其次简要分析氚系统功能和多重包容安全措施,划分出控制氚迁移的多重包容边界和流体回路边界,确定氚释放为PHA的重点。最后编制出PHA表,分析造成氚危险性释放的原因和危害后果,并给出预防措施。在PHA的基础上,列出应进行深入分析的几种典型事故。对氚设施和其他各种核设施安全水平的评估方法进行初步研究,对影响氚设施安全水平的几类重要因素进行分析,给出核设施安全水平的计算方法。     

Environmental and Safety Aspects of Using Tritium in Fusion

Today most current fusion research and development activity is based on the expectation that the D–T reaction will be used for the first generation of fusion reactors. This mixture is the premier candidate for a fusion fuel on account of its outstanding energy gain. Fusion reactors will produce neither the problematic emissions now experienced from fossil-fuel-burning power plants nor the long-lived fission products and transuranic elements resulting from fission reactors. Even though, tritium is a radioactive isotope of hydrogen with 12.32 years half-life that exposes beta radiation. Although its specific activity is relatively weak but because of its gaseous state, it can leak easily from its container and contaminate its surrounding. In order to improving safety and reliability of fusion reactors, research groups jointly investigate radiation hazards resulting from the release of tritium and activation products during normal operations as well as accidental conditions. In this paper, some of the most significant safety and environmental aspects of tritium is briefly discussed.     

产氚聚变堆概念设计系统研究

提出了基于球环类型的先进氚生产堆概念设计,它是聚变能发展的中间应用。与传统托卡马克氚生产堆不同,设计中利用了球形环的先进等离子体物理性能和紧凑的结构特征,并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,达到年生产氚1000 g的目标,相应的堆利用因子为40％。在2D中子学计算的基础上提出了较为完整的初步概念设计。逐项进行了分析,同时对设计的风险、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和重要的参考价值。     

Tritium Breeding Capability of Heliotron-H Fusion Reactor Blankets

Preparation and operation procedures of chromatographic column for hydrogen isotope separation have been examined. The best separation of isotopic molecular hydrogen was obtained when the stationary phase was activated at 230°C for 16 h and subsequently deactivated with CO₂ at ?7°C. The technique has been applied to analyzing commercially available tritium gases. Protium tritide (HT), DT, and tritiated-methane and -ethane were observed as impurities in all three samples analyzed. It was experimentally confirmed that most of the contaminant protium in the tritium gas came from the inner-surface of the storage vessel.