中国液态锂铅包层材料研究进展

液态锂铅包层是国际上普遍关注和最有发展潜力的聚变堆包层概念设计之一,而包层材料是液态锂铅包层的核心问题之一.目前,液态锂铅包层普遍选用低活化铁素体/马氏体钢(RAFM钢)作为结构材料,液态锂铅作为中子倍增剂及氚增殖剂.另外,部分设计采用了耐高温、电绝缘流道插件作为功能材料,以降低磁流体动力学效应及提高冷却剂出口温度(高于700℃).为适应液态包层的发展需求,中国科学院等离子体物理研究所FDS团队联合国内外相关研究单位,进行了具有中国自主知识产权的中国低活化马氏体钢(CLAM钢)及液态锂铅包层功能材料研发,并开展了锂铅热对流及强迫对流回路的设计、研制及腐蚀实验研究,以研究液态金属锂铅的流动特性及其与结构和功能材料的相容性.同时建立了聚变堆材料数据库平台,为促进中国聚变堆液态包层及材料技术的研究和发展提供数据支持.     

中国液态金属锂铅实验回路DRAGON-Ⅳ MHD实验段设计与分析

锂铅实验回路是研究液态金属锂铅包层关键技术问题的必备装置,MHD(Magneto Hydro Dynamic)实验是探究液态金属在聚变强磁场环境下流动特性的重要环节.本文结合聚变堆锂铅包层设计特点和实际的工程可行性,设计了中国高温多功能液态金属锂铅实验回路DRAGON-Ⅳ中的MHD实验段.初步结果表明MHD实验段设计方案的合理性,能够满足回路的设计原则和要求.     

ITER中国液态锂铅实验包层模块设计研究与实验策略

在广泛调研和深入分析国际聚变堆包层发展状况的基础上,根据液态锂铅包层一般特点和中国发展的系列液态锂铅包层概念设计,提出了一个具有演示氦气单冷却剂和氦气/锂铅双冷却剂包层技术的双功能包层模块实验系统方案,对其性能进行了分析研究,作为中国向ITER实验包层工作组(TBWG)提交的液态包层实验模块最终设计描述文件的内容框架。总结了该工作主要内容,包括基本设计思想和方案描述、性能分析概况、对辅助系统的要求和实验策略与关键技术等。     

聚变堆液态金属锂铅包层多功能涂层研发

液态金属锂铅包层是目前国际上聚变堆包层设计研究的主要方案之一,结构材料表面制备涂层是降低锂铅包层中的氚渗透率、液态锂铅腐蚀及磁流体动力学(MHD)效应的重要技术之一.本文主要从涂层材料及其制备工艺两个方面重点介绍了国内外在液态锂铅包层涂层材料研发方面的进展概况,并对涂层技术发展进行了展望,最后提出了中国发展液态锂铅包层涂层的规划建议.     

中国液态锂铅热对流实验回路DRAGON-Ⅰ温度场和速度场数值模拟分析

锂铅实验回路是研究聚变堆液态金属锂铅包层关键科学技术问题的必备装置,其锂铅温度场和速度场分布对分析相关实验结果是必需的。DRAGON-I是中国第一座热对流锂铅实验回路,采用FLU-ENT软件对该回路中液态锂铅的温度场和速度场进行了二维和三维模拟分析。结果表明:回路内锂铅温度场分布符合回路设计时的温度变化要求,并在回路管径截面上变化很小;锂铅速度除了在回路上端两个直拐角处波动较大外,其余管道直段波动非常小,平均速度约0.14 m/s。     

中国低活化马氏体钢CLAM在液态锂铅中腐蚀的初步实验研究

液态金属锂铅包层是最具发展潜力的聚变堆包层之一,其首选结构材料为低活化铁素体/马氏体钢,而它与液态锂铅的相容性是聚变堆材料研究领域的关键问题之一.本文介绍中国低活化马氏体钢CLAM在液态金属锂铅回路DRAGON-1热对流工况下的实验情况及500 h 480 ℃下初步腐蚀实验结果,并与同样工况下316L奥氏体钢腐蚀结果进行了对比分析.结果显示CLAM钢与液态锂铅的相容性优于316L钢.     

基于D-T聚变中子源的双功能液态铅锂包层中子学实验

中国双功能锂铅包层(Dual Functional Lithium-Lead,DFLL)是由中国科学院合肥物质科学研究院核能安全技术研究所设计的用于聚变反应堆的液态包层.由于聚变反应堆氚增殖包层的设计高度依赖于中子计算,为验证DFLL包层设计中所使用的核数据库和仿真软件,建立了DFLL包层实验模块,并基于D-T聚变中子...     

方管内液态铅锂流动MHD压降初步测量与分析

磁流体动力学(MHD)实验是研究聚变堆包层内液态铅锂在强磁场环境下流动特性的重要方法。本文基于中国多功能液态铅锂实验回路DRAGON-IV开展方管MHD压降测量实验,在磁场强度为1.88T和最大铅锂流量为0.36m~3/h时,方管实验段均匀磁场区压降为13.27kPa,与理论值偏差为2.30kPa,并对导致实验误差的因素进行了分析。     

CLAM钢基体上大气等离子体喷涂制备氧化铝涂层工艺研究

液态金属锂铅包层是目前国际上聚变堆包层设计研究的主流方案之一,但其仍面临高氚渗透率、液态锂铅对包层结构材料的腐蚀以及锂铅流动引起的磁流体动力学MHD效应等问题,而包层结构材料表面加覆涂层是解决上述问题的关键技术之一。本实验尝试利用大气等离子体喷涂(APS)工艺在中国低活化马氏体CLAM钢上制备多功能氧化铝涂层,试验结果表明:涂层与基底具有较好的结合强度,平均值为31.7 MPa;涂层电阻率为5.26×109~1.54×1010Ω.m;并呈现较高的显微硬度和致密度。本工作可为未来聚变堆液态锂铅包层涂层制备提供理论依据和技术储备。     

聚变发电反应堆双冷液态锂铅包层模块结构设计与分析

给出聚变发电反应堆FDS Ⅱ模块式液态锂铅包层(DLL)结构方案,以低活化马氏体(RAFM)钢为结构材料,采用液态金属LiPb作为增殖材料和冷却剂,使用碳化硅流道插件作为电绝缘和热绝缘。包层的设计特点体现在:从增殖区、冷却剂流道、屏蔽包层、母管、机械连接、维修装配等几个方面全局考虑包层设计,结构布置完整;独有的“”形隔板设计使氦气冷却回路容易实现,增殖流道简单,可简化制造工艺,提高可靠性。同其他液态锂铅包层相比,DLL包层在冷却剂系统、制造、装配上可成就较高的可行性。     

液态锂铅回路冷阱过滤芯捕集效率数值分析

液态锂铅纯化技术是聚变堆锂铅包层关键技术之一,冷阱是锂铅在线纯化的常用装置。鉴于冷阱过滤芯捕集杂质的效率难以在线测量,本文利用Fluent中离散相模型(DPM模型)对某些影响过滤芯捕集效率因素进行了数值模拟,得出了三种规格单层过滤芯周围的离散相浓度场分布及捕集效率。结果显示:在考虑杂质结晶析出速度的前提下,增大流体速度和降低过滤芯孔隙率可以提高过滤芯对杂质的捕集效率,且速度增大到一定数值之后,过滤芯对杂质捕集的效率增速减缓。相关结果可为冷阱过滤芯的的优化设计及液态金属流速的确定提供理论依据与参考。     

Preliminary numerical analysis of MHD flow of DFLL-TBM on the single coolant stage

The liquid lithium–lead (PbLi) breeder blanket concept has been explored extensively due to their potential attractiveness. To check and validate the feasibility, the China dual-functional lithium lead test blanket module (DFLL-TBM) system, which is designated to demonstrate the integrated technologies of both He single coolant (SLL: single-cooled lithium lead) and He–LiPb dual-coolant (DLL: dual-cooled lithium lead) blankets, is proposed for test in ITER. One of the key feasibility issues is the impact of liquid metal MHD effect which will influence the pressure drop, flow distribution, and heat transfer in a DFLL-TBM.To reduce MHD effect, an electrically insulating coating is applied onto the inner surface of the flow channel for single coolant blanket. In this work, a preliminary numerical study of MHD flows in a simplified DFLL-TBM model on the single coolant stage has been carried out to assess the performance of such a concept with regard to the above mentioned MHD problems and constraints. The flow distribution and MHD pressure drop of LiPb flow in the SLL stage TBM are analyzed.     

Preliminary experiments on the corrosion of CLAM steel in flowing eutectic LiPb

China low activation martensitic steel (CLAM), which is one of the RAFMs and under development in ASIPP, is considered as the primary candidate structural material for blankets in FDS series fusion reactors and will be in contact with the liquid breeder LiPb, so its corrosion behavior is of significance for its successful application. In the thermal convection LiPb loop DRAGON-I, the corrosion experiment of CLAM in flowing LiPb at 480 °C was performed up to about 2500 h to evaluate the corrosion behavior of CLAM in LiPb. The exposed samples were observed and analyzed, respectively, by metallography and scanning electron microscopy (SEM) with energy dispersive X-ray (EDX) spectroscopy after 500, 1000, 1500, 2000 and 2500 h corrosion experiments, respectively. The results show that the corrosive attack started with the dissolution of the passivated oxides formed during the heat treatment on the surface of CLAM steel. Then the corrosive attack continued further into the steel matrix. The weight loss was about 0.38 mg/cm² after 2500 h exposure.     

Status of gas-cooled fast breeder reactor programs

Most gas-cooled fast breeder reactor (GCFR) programs in Europe and the US are now coordinated and focused on a 300 MW(e) GCFR demonstration plant program. Except for venting and artificial surface roughening, GCFR fuel is similar to liquid metal fast breeder reactor (LMFBR) fuel and operates under nearly identical conditions. The primary helium system is integrated within a PCRV like all large gas-cooled thermal reactors, with three main loops and three auxiliary loops. Design and safety studies and various experiments, including heat transfer, irradiation, and critical experiments, indicate that most feasibility questions have been answered and a demonstration plant could be in operation within 12 years. This could be followed in the mid-1990s by a large-size GCFR with a doubling time of about 10 years fueled by (UO₂---PuO₂) and producing either ²³³U in thorium blankets as fuel for advanced converters or plutonium in depleted uranium blankets.     

Neutronics study for the installation of the LBVM in the medium flux area of IFMIF

IFMIF (International Fusion Materials Irradiation Facility) will be a fusion dedicated facility producing a large amount of neutrons with the appropriate energy spectrum to test materials and subcomponents for DEMO and future Fusion Power Plants.While the high flux area of IFMIF will be devoted to reduced activation structural materials for first wall and blanket, the medium flux area will be dedicated to functional materials for breeder blankets. In particular, the Liquid Breeder Validation Module (LBVM), will host experiments related with functional materials for liquid breeder blankets. Since IFMIF neutron spectra have been intended to fit the most irradiated areas of a fusion reactor in the high flux area, the irradiation conditions in the LBVM placed in the medium flux area of IFMIF have been assessed. The effect of some neutron shifter/reflector components to optimize the neutron spectra have been evaluated in order to find out the proper irradiation conditions for functional materials for liquid breeder blankets.Therefore, the objective of this report is to summarize the neutronic calculations developed to evaluate the viability of IFMIF neutron source to perform relevant irradiation experiments on functional materials for liquid breeder blanket concept for future nuclear fusion power reactors (ITER, DEMO). The irradiation parameters evaluated for this purpose are: the tritium production for liquid breeder material (Pb–17Li) and the damage dose (dpa) and gas production to damage dose ratios for Al₂O₃ and SiC functional materials.The main conclusion is that, it is possible to perform relevant irradiation experiments on functional materials for liquid breeder blanket concept for the future nuclear fusion reactor DEMO. Nevertless, the use of some shifter components will be needed to optimize some irradiation parameters.     

液态锂铅合金的氚释放行为

为了完成聚变堆液态锂铅包层鼓泡提氚系统的工程设计和建造,以金属与氢的作用理论为基础,建立了氚从液态锂铅中的动力学释放行为的数学模型.计算和分析了温度、饱和器氚分压、氦流量对解吸器顶部气相中的氚分压的影响以及氚在液态锂铅中的传质系数、解吸率和吸附率.结果表明:在633～723 K的解吸温度范围内,氚从液态锂铅到气相的整个释放过程虽然包含了氚在熔融合金气泡中的扩散与对流、氚通过与气-液界面相连合金层的扩散、在界而发生的氚原子重组多相反应、氚通过气相边界层的扩散和气相中氚的扩散与对流5个子过程,但起决定作用的是氚在合金内的扩散和气.液界面的多相反应重组,其他子过程意义不大.     

Neutronics Optimization of Tritium Breeding Blanket for the FDS

Neutronics optimization calculations have been performed for the tritium breeding blankets with solid ceramic breeder Li2O and liquid eutectic breeder Li17Pb83,respectively,based on a 2-D geometrical configuration using the Monte Carlo neutron-photon transport code MCNP/4B.The effects of beryllium.^6Li enrichment and various structural materials on Tritium Breeding Ratio have been systematically analyzed.     

Tritium permeation experiment at IFMIF Medium Flux Test Module

Tritium release experiments using different breeding material candidates are planned for the medium flux region of the IFMIF Test Cell. Nowadays, only ceramic breeder materials have been suggested to be tested in the Tritium Release Module located in the Medium Flux Test Module of IFMIF. Liquid breeder blankets are very promising and for that reason, several concepts will be tested in ITER. One of the main problems concerning the liquid blankets is the permeation of the generated tritium in the breeder throughout the walls. Since tritium permeation is highly influenced by irradiation conditions, IFMIF is a suitable scenario to perform tritium permeation related experiments.In this paper, a preliminary design of a tritium permeation experiment for the Medium Flux Test Module of IFMIF is proposed, in order to contribute to the progress of the liquid breeder blanket concept validation.The conceptual design of the capsule in which the experiment will be performed is carried out, taking into consideration the experiment necessities and its implementation in the Tritium Release Module. In addition to this, some thermal hydraulic calculations have been performed to evaluate the thermal behaviour of the irradiation capsule.     

The fusion breeder

The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the U.S. fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the U.S. fusion program and the U.S. nuclear energy program. There is wide agreement that many approaches will work and will produce fuel for five equal-sized LWRs, and some approach as many as 20 LWRs at electricity costs within 20% of those at today's price of uranium ($30/lb of U₃O₈). The blankets designed to suppress fissioning, called symbiotes, fusion fuel factories, or just fusion breeders, will have safety characteristics more like pure fusion reactors and will support as many as 15 equal power LWRs. The blankets designed to maximize fast fission of fertile material will have safety characteristics more like fission reactors and will support 5 LWRs. This author strongly recommends development of the fission suppressed blanket type, a point of view not agreed upon by everyone. There is, however, wide agreement that, to meet the market price for uranium which would result in LWR electricity within 20% of today's cost with either blanket type, fusion components can cost severalfold more than would be allowed for pure fusion to meet the goal of making electricity alone at 20% over today's fission costs. Also widely agreed is that the critical-path-item for the fusion breeder is fusion development itself; however, development of fusion breeder specific items (blankets, fuel cycle) should be started now in order to have the fusion breeder by the time the rise in uranium prices forces other more costly choices.     

Progress in compatibility experiments on lithium–lead with candidate structural materials for fusion in China

Liquid LiPb eutectic is one of the promising candidate tritium breeder materials for fusion reactors. This paper presents the progress in compatibility experiments with liquid LiPb achieved up to now in China for some candidate structural materials. The results showed that CLAM steel had good compatibility with flowing LiPb at 480 °C with the velocity of 0.08 m/s after 5000 h in DRAGON-I loop. On the other hand, after exposed in static LiPb at 700 °C for 500 h in a SiC crucible, the W and Mo specimens suffered much more weight loss compared with Nb specimen, and a thin reaction product layer was visible on the surface of all the three refractory metals. Preliminary analysis on SiC_f/SiC composite specimens indicated that there was no penetration of LiPb and no reaction products on the surface with CVD SiC coating, which showed SiCf/SiC composite were stable and compatible with static LiPb under 700 °C after 500 h exposure.