Experimental design of tritium extraction loop from lead lithium eutectic

Effective tritium breeding achievable in Test Blanket Module (TBM) is a major issue for sustainable fusion energy program. Equally important is tritium extraction to recover and recycle tritium back to fusion reactor. Tritium extraction from lead lithium is much more complicated than from purge gas due to low tritium extraction efficiency in transfer step to gas phase and the limitations imposed on space and lead lithium inventory in port cell. Earlier investigations do suggest the preference of packed columns over bubble columns. Theoretical models based on axial dispersion plug flow in liquid and gas proposed for bubble columns and packed columns are reinvestigated for different boundary conditions.This paper highlights the critical issues of experimental design based on tritium extraction efficiency and its impact on recovery loop. Steady state closed loop for absorption and stripping of hydrogen isotopes using inert gas is designed along with the associated auxiliaries.     

Status of Indian LLCB TBM program and R&D activities

The Indian LLCB TBM, currently under development, will be tested from the first phase of ITER operation (H–H phase) in one-half of the ITER port no-2. The present LLCB TBM design has been optimized based on the neutronic as well as thermal hydraulic analysis results. LLCB TBM R&D activities are primarily focused on (i) development of technologies related to various process systems such as Helium, Pb–Li liquid metal and tritium, (ii) development and qualification of blanket materials viz., structural material (IN-RAFMS), tritium breeding materials (Pb–Li, and Li₂TiO₃), (iii) development and qualification of fabrication technologies for TBM system. The present status of LLCB TBM design activities as well as the progress made in major R&D areas is presented in this paper.     

厚度和湿度对氚在一些橡胶和塑料中的渗透率的影响

为了研究材料结构的完整性及其湿度对氚的渗透率的影响,我们在过去工作的基础上,进行了材料厚度和湿度对氚在一些橡胶和塑料中的渗透率的实验测量工作。     

Development of a potential based code for MHD analysis of LLCB TBM

A two dimensional solver is developed for MHD flows with low magnetic Reynolds’ number based on the electrostatic potential formulation for the Lorentz forces and current densities which will be used to calculate the MHD pressure drop inside the channels of liquid breeder based Test Blanket Modules (TBMs). The flow geometry is assumed to be rectangular and perpendicular to the flow direction, with flow and electrostatic potential variations along the flow direction neglected. A structured, non-uniform, collocated grid is used in the calculation, where the velocity (u), pressure (p) and electrostatic potential (?) are calculated at the cell centers, whereas the current densities are calculated at the cell faces. Special relaxation techniques are employed in calculating the electrostatic potential for ensuring the divergence-free condition for current density. The code is benchmarked over a square channel for various Hartmann numbers up to 25,000 with and without insulation coatings by (i) comparing the pressure drop with the approximate analytical results found in literature and (ii) comparing the pressure drop with the ones obtained in our previous calculations based on the induction formulation for the electromagnetic part. Finally the code is used to determine the MHD pressure drop in case of LLCB TBM. The code gives similar results as obtained by us in our previous calculations based on the induction formulation. However, the convergence is much faster in case of potential based code.     

Tritium management and safety issues in ITER and DEMO breeding blankets

Safe, reliable and efficient tritium management in the breeder blanket faces unique technological challenges. Beside the tritium recovery efficiency in the tritium extraction and coolant purification systems, the tritium tracking accuracy between the inner and outer fuel cycle shall also be demonstrated. Furthermore, it is self-evident that safe handling and confinement of tritium need to be stringently assured to evolve fusion as a reliable technique. The present paper gives an overview of tritium management in breeder blankets. After a short introduction into the tritium fuel cycle and blanket basics, open tritium issues are discussed, thereby focusing on tritium extraction from blanket, coolant detritiation and tritium analytics and accountancy, necessary for accurate and reliable processing as well as for book-keeping.     

Adsorption capacity of hydrogen isotopes on mordenite

In a fusion reactor system, a monitoring of hydrogen isotopes including tritium is necessary for the safety of system control and operation. A gas chromatography using a cryogenic separation column is one of the methods for hydrogen isotope analysis. Synthesis zeolite such as molecular sieve 5A (CaA) is a candidate material of the separation column, and its property varies by the ratio of silica to alumina, the kinds of cation and so on. If the factor affected the hydrogen adsorption property of the synthesis zeolite is clarified, it may lead to the development of the new zeolite optimized to the separation column. So, in this work, adsorption capacity of hydrogen (H₂) and deuterium (D₂) for mordenite (MOR) and NaY type zeolite (NaY) were investigated at various temperatures, and were compared with CaA. The amount of adsorption per unit weight of MOR was larger than that of CaA, and that of NaY was smaller than that of CaA. The adsorption isotherms were expressed by sum of two Langmuir equations, and the Langmuir coefficients of H₂ and D₂ were proposed. Furthermore, the Langmuir coefficients of HD, HT, DT and T₂ were estimated by the reduced mass. The correlation between the adsorption properties and the physical parameters of the zeolite were not confirmed.     

Engineering studies for integration of the test blanket module (TBM) systems inside an ITER equatorial port plug

The European test blanket module (EU-TBM), first prototype of the breeding blanket concepts under development for the future DEMO power plant to produce the tritium, will be developed to be tested in three equatorial ports of ITER dedicated to this. The CEA Cadarache under the contract of Association EURATOM/CEA and in close relation with Association EURATOM/HAS works on the integration of the EU-TBM inside ITER tokamak.The installation of the TBM into the vacuum vessel is made with the help of a port plug, constituted with two components: the Shield module and the Port-Plug frame. The Shield module provides the neutron shielding inside the Port-Plug frame, which maintains in cantilever position the TBM and its shield module and closes the vacuum vessel port.This paper will describe the EU-TBM design and integration activities on the cooled shield module and on its interface with the TBM component. A particular attention, in term of thermal and mechanical studies, is dedicated to the design of the shield and test blanket module attachment, and also to the shield design and its internal cooling system.     

氚在注入铝,氧和钼,铬离子的铀中的扩散渗透行为研究

一、引言铀与氚有较强的化学反应,这种反应随氚的压力和温度的增加而加剧。当氚在金属(?)中的浓度低于铀对氚的固溶极限时,氚与铀发生非常缓慢的氚化反应,这时,氚从高浓度端     

低温精馏氢同位素分离影响因素研究

本文系统研究了低温精馏氢同位素分离中总板数、进料位置、回流比、采出率等操作条件对系统分离性能的影响,获得了精馏柱上的浓度和温度分布.随着进液位置向底端移动,再沸器和冷凝器中HD浓度均减小;随着回流比的增大,再沸器和冷凝器中HD浓度均减小;顶端采出率增大,再沸器中HD浓度明显增大;在相同的总板数下,H2/HD和D2/DT两个体系的分离特性明显有差别.     

D2/DT低温精馏分离动态模拟

本文建立了氢同位素低温精馏分离的动态模型,利用该模型针对D2/DT体系计算获得了精馏柱上DT浓度的动态变化和空间分布;研究了再沸器滞留量、回流比等参数对系统分离性能的影响.回流比的提高可以显著地提高脱氚率;再沸器中DT浓度不但与滞液量存在显著的依赖关系,而且随时间增长而增长.在理论板数为80时,塔顶与塔底DT浓度相差约3个数量级,分离效果明显.     

双组分体系低温蒸馏氢同位素分离理论计算

建立了单根柱的低温蒸馏计算模型,对H2/HD,H2/HT和D2/DT三个双组分体系进行了计算。给出了温度和各组分浓度在柱上的空间分布,结果较好地再现了易挥发组分在柱高端浓集,而难挥发组分在柱底端浓集的分离特征。三个体系以H2/HT最易分离,H2/HD次之,D2/DT最差。对于文中研究的体系,当理论板数为80时,D2/DT不能达到脱氚率99%的分离要求。     

疏水催化剂用于HD/H2O同位素交换的性能研究

系统地研究了气体线速度、温度等对疏水催化剂传质系数的影响.气体线速度在2-9 cm·s-1范围内,催化剂传质系数呈线性增长.气体线速度在20-90 cm·s-1范围内,催化剂传质系数在40-60mol·m-3·s-1范围内,通过提高气体线速度来提高传质系数是一有效途径.降低进液位置后发现浓缩段传质系数在高气速下显著下降,从实验上证实了蒸汽在催化交换过程中具有极其重要的作用.操作温度宜在45℃左右.     

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.     

低温精馏氢同位素分离技术及其应用

低温精馏是大规模分离氢同位素的有效手段,本文对该工艺涉及的气体纯化、制冷与真空、测量与控制、安全防护等关键技术进行简要介绍。对该工艺在重水生产、重水升级和除氚、聚变堆氘氚燃料循环、武器用氚等领域的国外应用做了回顾。分析了国内对该技术的需求,提出今后开展的研究方向。     

氚在一些弹性材料和聚合材料中渗透率的测量

在制造氚容器、氚操作系统(如手套箱)的材料中,弹性材料和聚合材料是十分重要,而且几乎是不可缺少的材料。例如,各种橡胶、塑料和密封用的粘接材料都广泛应用于制造手套、垫圈、维修用帐篷以及氚防护服等。氚在这些材料中的渗透率一般都比在金属中     

Mathematical models for tritium permeation analysis in liquid metal flows with helium bubbles

In LIBRETTO-2 test, evidence was obtained that helium bubbles nucleated and grew in the neutron irradiated PbLi probes. If such phenomenon occurs inside liquid metal (LM) breeding blanket channels, the study of its effect on tritium permeation and heat transfer in the near wall region will acquire utmost importance. The T4F research group has developed in the past a nucleation, growth and transport model for helium bubbles in LM flows, as well as a tritium transport model in such a multi-fluid system. In the present study, we are focused on the near-wall region analysis in order to obtain a wall function that allow reproducing the tritium permeation with coarse meshes and, hence, reduce the computational time. First, we perform some detailed CFD simulations of the near-wall region where bubbles might be attached. In these simulations, tritium diffusion processes as well as tritium recombination and dissociation are modelled. The analysis of such simulations allows us to further understand the complex phenomena and justify the use of simplified models. As a result, a new model for tritium transport across a LM–solid interface partially covered by helium bubbles is developed, implemented and validated. This simplified model can be seen as a wall function for the CFD simulation which substantially reduces computational time.