带热浸铝涂层MANET Ⅱ马氏体钢的氢渗透性能研究

在300～450℃温度范围内,分别在氢气相和液态铅锂合金相中,开展了带热浸铝涂层MANET Ⅱ马氏体钢的氢渗透性能研究.结果表明,实验所得到的氢渗透率减低因子(PRF),在气相中为620～263,在液态铅锂合金相中为45～30.但仍然没有满足DEMO聚变堆的设计要求.涂层的自修复效应在400 ℃以上是明显和有效的.从渗透通量与样品上游氢压的关系来看,涂层使得表面效应对渗透过程的影响很大.在上游小流量及在液态铅锂合金中鼓泡充氢可以导致渗透通量升高.扫描电镜和能谱分析(SEM-EDS)的结果表明,样品表面被液态铅锂合金所浸润的部分出现微裂纹,而未浸润部分没有出现微裂纹.微裂纹很肤浅,仅仅影响涂层的最外表面薄层,没有贯穿整个渗铝层而到达基体.未浸润表面的Al/O原子比约为2/3,浸润表面约为1/1,表明液态铅锂合金对渗铝层表面的Al2O3薄层造成了损伤.总的看来,造成氢渗透阻挡层性能退化的原因,是涂层外表面与液态铅锂合金相互作用,以及涂层在升、降温过程中产生热应力释放.     

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.     

ITER中国液态锂铅实验包层模块氚渗透分析

对ITER中国液态锂铅实验包层模块的氚渗透途径进行了初步分析,并建立了氚渗透模型;在确保环境安全的前提下,通过计算LiPb中的氚分压分析了氚渗透量及氚总量的分配情况;在此基础上通过改变进入氚提取系统中LiPb比例(F)和涂层氚渗透减少因子(TPRF)对氚提取及渗透的影响做了灵敏性分析.     

基于扩散界面法的液态铅铋合金中气泡上升行为模拟

基于扩散界面法,对单个氮气气泡在液态铅铋合金内从静止到充分发展整个过程中的动力学行为进行数值模拟,得到气泡形变特性和气泡上升速度随时间的变化关系,将模拟结果与Grace经验关系图对比,发现模拟得到的气泡形变结果在Grace经验关系图中均可找到且很好地吻合,从而验证了扩散界面法在模拟液态铅铋合金中气泡上升行为的可行性和准确性。同时基于界面扩散法的模拟,对比了5种不同初始直径的氮气泡在液态铅铋合金中的上升行为,发现初始直径较小的气泡在上升过程中扰动会更剧烈,初始直径较大的气泡在上升过程中易发生分裂现象。     

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.     

ITER中国液态锂铅实验包层模块氚提取系统设计

ITER中国液态锂铅实验包层模块氚提取系统(TES)是通过含0.1%H2的低压氦吹洗气流,在鼓泡器中将液态锂铅内产生的氚交换和载带出来,进入同位素分离系统连接进行氚提取.给出了该系统总体参数、工艺流程、辅助设施等设计.     

基于扩散界面法的液态LBE中单个弹状气泡上升行为CFD研究

利用扩散界面法对液态铅铋合金中弹状气泡在垂直管中的上升过程进行数值模拟研究,用来验证扩散界面法在模拟液态铅铋合金(LBE)中弹状气泡上升行为的准确性和可行性,并对模拟结果进行进一步分析,来解释弹状气泡在LBE中的上升行为。数值模拟得到不同液体流速情况下弹状气泡在上升过程中的形态和速度变化,数值模拟结果与文献中的经验关系式以及实验中的结果都吻合良好,证明了扩散界面法在模拟液态铅铋合金中弹状气泡上升行为的准确性和可行性,为液态金属和弹状气泡的两相流提供了一种新方法。对数值模拟结果进行进一步分析,发现气泡尾部形态变化较大,会在尾部两端分裂出子气泡,尾部附近流场产生旋涡,受到强烈干扰而出现湍流,对于提高换热效率起到积极作用。气泡上升对尾部区域产生影响的最远距离约80 mm,为连续气泡的注入提供了气泡间距参考值。     

液态金属内单个气泡上升行为的MPS法数值模拟

液态金属冷却核反应堆采用气泡泵的概念设计来提升堆芯自然循环能力。液态金属内气液两相流动特征将直接影响核反应系统一回路的自然循环能力及堆芯安全。本研究通过采用移动粒子半隐式（MPS）方法,对液态金属中单个上升气泡的气泡动力学行为进行数值模拟。分析了铅铋合金中3种初始直径不同的单个氮气泡在上升过程中的气泡形状和速度的变化趋势;对比了初始直径相同的单个氮气泡在液钾、液钠、铅铋合金、钾钠合金和锂铅合金5种液态金属中的上升行为;同时将模拟得到的气泡形状与Grace经验关系图进行了对比,验证了MPS方法数值模拟结果的正确性。     

液态锂铅合金中316L不锈钢的静态腐蚀行为

采用挂片法、失重法和金相分析,开展了结构材料316L不锈钢在液态锂铅(LiPb)合金中静态腐蚀行为的研究.研究结果表明:316L不锈钢中的组分元素,在液态LiPb合金中发生了溶解和质量迁移,这是导致材料腐蚀的主要原因,而温度和合金中的氧含量是影响静态腐蚀行为最重要的参数.     

液态锂铅合金中测氢传感器的探头材料

氢同位素在液态锂铅合金中的溶解行为是聚变堆液态包层提氚系统设计的重要参考指标。为了解决液态锂铅合金中氢含量测定的技术难题,研发测氢传感器,开展了以Al2O3、SiC、SiO2-Cr2O3、TiC为候选的传感器探头材料的选型研究。研究结果表明:在仅考虑测量平衡时间、合金熔融温度和热冲击的前提条件下,SiC探头材料是最合适的选择,具有测量直观准确、高热稳定性、抗化学腐蚀性能和耐热冲击性能良好的特点;而Al2O3和SiO2-Cr2O3在高温锂铅环境中易腐蚀;TiC高温下易氧化,但在873K以下时的测量值比SiC更接近理论饱和值。     

Tritium breeding control within liquid metal blankets

A key requirement for DEMO is the on-site breeding of tritium. In order to do this, a robust control system must be employed to ensure enough tritium is being bred to sustain the fusion reactor, whilst not breeding an amount which would exceed the plant's tritium inventory license. A tritium breeding method which is cost effective and reduces radioactive waste for disposal is that of the liquid metal breeder such as those based around LiPB and FLiBe. This paper focuses on the modeling of a simplified fusion reactor design with a LiPb blanket with linked radiation transport, nuclide burn-up and control theory. Two simple models were simulated using the FATI code which incorporated a PID (proportional integral derivative) controller that adjusted the Li6/Li7 ratio in order to increase/decrease tritium production based on the difference between the measured excess tritium inventory and the desired excess inventory. The modelling has initially demonstrated that a linear PID controller has the capability to manage tritium production within a LiPb liquid blanket.     

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.     

Design and analysis on tritium system of multi-functional experimental fusion–fission hybrid reactor (FDS-MFX)

As early application of fusion technology, the fusion–fission hybrid systems/reactors could be used to transmute long-lived radioactive waste and produce fissile nuclear fuel. A fusion–fission hybrid reactor named FDS-MFX was designated for checking and validating the DEMO reactor blanket relevant technologies. The reactor design is based on easy-achieved plasma parameters extrapolated from the successful operation of tokamaks and the subcritical blanket is designed based on the well-developed technologies of fission reactors. In this contribution, a concept of the tritium system was designed for the FDS-MFX: the tritium was extracted from LiPb by the helium purge gas which contains a small amount of hydrogen gas, then the impurity gas was removed by cold trap, finally tritium was separated from hydrogen isotope by the cryogenic distillation and supply to reactor core. On the basis of data obtained by present design and experimental research, the system parameters were presented and discussed in detail. The results preliminarily demonstrated the engineering feasibility of the design.     

Neutronics and pumping power analyses on the Tokamak reactor for the fusion-biomass hybrid concept

The authors aim to develop a fusion-biomass combined plant concept with a small power fusion reactor. A concern for the small power reactor is the coolant pumping power which may significantly decreases the apparent energy outcome. Thus pressure loss and corresponding pumping power were studied for a designed Tokamak reactor: GNOME. First, 3-D Monte-Carlo Neutron transport analysis for the reactor model with dual-coolant blankets was taken in order to simulate the tritium breeding ability and the distribution of nuclear heat. Considering calculated concentration of nuclear heat on the in-board blankets, pressure loss of the liquid LiPb at coolant pipes due to MHD and friction forces was analyzed as a function of fusion power. It was found that as the fusion power increases, the pressure loss and corresponding pumping power exponentially increase. Consequently, the proportion of the pumping power to the fusion power increases as the fusion power increases. In case of ～360 MW fusion power operation, pumping power required for in-board cooling pipes was estimated as ～1% of the fusion power.     

液态锂锡合金氚增殖行为的理论分析

采用气-液两相界面模型和与时间有关的扩散理论及本征函数展开的方法,模拟了Li25Sn75合金的氚增殖行为.计算结果表明:在14 MeV能量下,天然Sn的(n,2n)反应宏观截面相对较小,只有1.5 b;7Li、6Li产氚随时间变化的规律与LiPb合金、Li2O介质是一致的;Li25Sn75合金对模型厚度比较敏感,随着厚度和6Li丰度的增加,氚增殖比(Tritium Breeding Ratio,TBR)保持上升的趋势.     

Preliminary thermal-hydraulic and safety analysis of China DFLL-TBM system

China has proposed the dual-functional lithium-lead (DFLL) tritium breeding blanket concept for testing in ITER as a test blanket module (TBM), to demonstrate the technologies of tritium self-sufficiency, high-grade heat extraction and efficient electricity production which are needed for DEMO and fusion power plant. Safety assessment of the TBM and its auxiliary system should be conducted to deal with ITER safety issues directly caused by the TBM system failure during the design process. In this work, three potential initial events (PIEs) – in-vessel loss of helium (He) coolant and ex-vessel loss of He coolant and loss of flow without scram (LOFWS) – were analyzed for the TBM system with a modified version of the RELAP5/MOD3 code containing liquid lithium-lead eutectic (LiPb). The code also comprised an empirical expression for MHD pressure drop relevant to three-dimensional (3D) effect, the Lubarsky–Kaufman convective heat transfer correlation for LiPb flow and the Gnielinski convective heat transfer correlation for He flow. Since both LiPb and He serve as TBM coolants, the LiPb and He ancillary cooling systems were modeled to investigate the thermal-hydraulic characteristic of the TBM system and its influence on ITER safety under those accident conditions. The TBM components and the coolants flow within the TBM were simulated with one-dimensional heat structures and their associated hydrodynamic components. ITER enclosures including vacuum vessel (VV), port cell and TCWS vault were also covered in the model for accident analyses. Through this best estimate approach, the calculation indicated that the current design of DFLL-TBM and its auxiliary system meets the thermal-hydraulic and safety requirements from ITER.     

聚变-裂变混合堆水冷包层中子物理性能研究

研究直接应用国际热核聚变实验堆（ITER）规模的聚变堆作为中子驱动源,采用天然铀为初装核燃料,并采用现有压水堆核电厂成熟的轻水慢化和冷却技术,设计聚变-裂变混合堆裂变及产氚包层的技术可行性。应用MCNP与Origen2相耦合的程序进行计算分析,研究不同核燃料对包层有效增殖系数、氚增殖比、能量放大系数和外中子源效率等中子物理性能的影响。计算分析结果显示,现有核电厂广泛使用的UO₂核燃料以及下一代裂变堆推荐采用的UC、UN和U₉₀Zr₁₀等高性能陶瓷及合金核燃料作为水冷包层的核燃料,都能满足以产能发电为设计目标的新型聚变 裂变混合堆能量放大倍数的设计要求,但只有UC和U₉₀Zr₁₀燃料同时满足聚变燃料氚的生产与消耗自持的要求。研究结果对进一步研发满足未来核能可持续发展的新型聚变-裂变混合堆技术具有潜在参考价值。     

Toward the ultimate goal of tritium self-sufficiency: Technical issues and requirements imposed on ARIES advanced power plants

Due to the lack of external tritium sources, all fusion power plants must demonstrate a closed tritium fuel cycle. The tritium breeding ratio (TBR) must exceed unity by a certain margin. The key question is: how large is this margin and how high should the calculated TBR be? The TBR requirement is design and breeder-dependent and evolves with time. At present, the ARIES requirement is 1.1 for the calculated overall TBR of LiPb systems. The Net TBR during plant operation could be around 1.01. The difference accounts for deficiencies in the design elements (nuclear data evaluation, neutronics code validation, and 3D modeling tools). Such a low Net TBR of 1.01 is potentially achievable in advanced designs employing advanced physics and technology. A dedicated R&D effort will reduce the difference between the calculated TBR and Net TBR. A generic breeding issue encountered in all fusion designs is whether any fusion design will over-breed or under-breed during plant operation. To achieve the required Net TBR with sufficient precision, an online control of tritium breeding is highly recommended for all fusion designs. This can easily be achieved for liquid breeders through online adjustment of Li enrichment.