Measurements of time-dependent liquid-metal magnetohydrodynamic flows in a flat rectangular duct

In the helium-cooled lead lithium (HCLL) blanket, which has been chosen as a reference concept for a liquid-metal breeding blanket to be tested in ITER, the heat is removed by helium cooled plates aligned with the strong toroidal magnetic field that confines the fusion plasma. The liquid breeder lead lithium circulates through gaps of rectangular cross-section between the cooling plates to transport the generated tritium towards external extraction facilities. Under the action of the strong magnetic field, liquid metal flows in conducting rectangular ducts exhibit jet-like velocity profiles in the thin boundary layers near the side walls, which are parallel to the magnetic field like the cooling plates in HCLL blankets. The velocity in these side layers may exceed several times the mean velocity in the duct and it is known that these layers become unstable for sufficiently high Reynolds numbers. The present paper summarizes experimental results for such unstable time-dependent flows in strong magnetic fields, which have been obtained in the MEKKA liquid metal laboratory of the Forschungszentrum Karlsruhe. In particular, spatial and temporal scales of perturbation patterns are identified. The results suggest that the flow between cooling plates in a HCLL blanket is laminar and stable. The observed time-dependent flow behavior appears at larger velocities so that the present results are more relevant for applications in dual coolant concepts where high-velocity jets have been predicted along side walls.     

Influence of helium cooling channels on magnetohydrodynamic flows in the HCLL blanket

One of the blanket concepts proposed to be tested in ITER as part of the test blanket module program of the European Union is the helium cooled lead lithium blanket design. In this configuration the so called breeder units are arranged in an array, separated by a stiffening grid, to form blanket modules. The deposited thermal energy is removed by helium flowing at high pressure and speed in channels integrated both in the walls and in cooling plates that subdivide the breeder units into flat ducts where the lead lithium circulates under the influence of the strong plasma confining magnetic field. This gives rise to magnetohydrodynamic (MHD) phenomena whose effects on flow distribution have to be investigated to evaluate the performance of the proposed design. The established MHD flow is affected by the presence of helium channels in cooling and stiffening plates that results in non-homogeneous wall conductance.In support to the conceptual study of a liquid metal blanket, numerical investigations of fully developed MHD flows in a central cross-section of breeder units have been performed, taking into account both the presence of helium channels in the walls and the multi-channel effects caused by the exchange of currents through walls separating different fluid domains.     

Finite element analysis of magnetohydrodynamics and its application to lithium blanket design of a fusion reactor

This paper presents the finite element analyses of magnetohydrodynamic (MHD) flow and its application to liquid lithium blanket design of a fusion reactor. In the preliminary analyses of MHD flow, we are concerned with both coupling and uncoupling problems between the magnetic and velocity fields. Next, the steady state liquid lithium flow in the blanket of fusion reactor is analysed with the practical simplification of the MHD equations for a typical fusion reactor. In addition, thermal and mechanical behaviors of the blanket are also discussed. Finally, we study the transient problem of MHD flow when the plasma current linearly drops in a short time, where we take account of the compressibility of liquid lithium because the phenomena should be considered as a kind of shock caused by electromagnetic forces, and the transient magnetic and electric fields are also solved with axisymmetric equations of vector potential as inputs for the analysis.     

Application of Numerical Simulation in the Design of HECLIC Blanket for CFETR

China Fusion Engineering Test Reactor is a new tokamak device which is proposed by China National Integration Design Group. The fusion power is 50–200 MW and its plasma major radius and plasma minor radius are 5.7 and 1.6 m. The helium cooled lithium ceramic (HECLIC) blanket, as a key component of the tokamak, has the basic function to provide tritium breeding and plasma limiter. The blanket also provides main thermal and nuclear shielding of the vacuum vessel and ex-vessel components such as magnetic coils during plasma operations. With the development of the numerical simulation technology, more and more design parameters can be obtained by this method. Numerical simulation has been used for design and optimization, because some parameters are very hard to obtain though theoretical calculation. In this study, the simulation methods are investigated for HECLIC blanket design. Besides, design flow of the blanket is discussed and related analysis is also introduced to improve the design.     

Status and prospect of the development of liquid lithium limiters for stellarotor TJ-II

Stellarator concept is considered as a promising approach for power fusion reactor development because it basically free from disruptions and other extreme thermal load events. However, the potential problem of impurity accumulation in stellarator plasma should be taken into account. Very promising results in density control, plasma reproducibility and confinement characteristics have been obtained with application of “lithiation” technology. The next step in the improvement of TJ-II Heliac plasma performance is the development and creation of two poloidal liquid lithium limiters (LL). Experimental possibilities, design, structural materials and main parameters of LL based on capillary-pore structure (CPS) filled with liquid lithium are considered. Understanding in hydrogen isotope interaction with liquid lithium surface is an important aspect of lithium technology development for fusion reactor application. Therefore study of deuterium sorption/desorption process on a lithium surface of LL is stipulated. The development of lithium CPS based devices decreasing intensity of plasma–wall interaction on a central “groove” of TJ-II vacuum camera is proposed as the further step in plasma performance improvement owing to decrease in impurity flux from the wall.     

A new facility for studying plasma interacting with flowing liquid lithium surface

A new facility to study plasmas interacting with flowing liquid lithium surface was designed and is constructing in Sichuan University. The integrated setup includes the liquid lithium circulating part and linear high density plasma generator. The circulating part is consisted of main loop, on-line monitor system, lithium purification system and temperature programmed desorption system. In our group a linear high density plasma generator was built in 2012. Three coils were mounted along the vessel to produce an axial magnetic field inside. The magnetic field strength is up to 0.45 T and work continuously. Experiments on plasmas interacting with free flowing liquid lithium surface will be performed.     

TCB混合堆Li自冷包层MHD流动设计

从磁流体动力学ＭＨＤ压降的物理原理出发，对ＴＣＢ商用混合堆Ｌｉ自冷包层的ＭＨＤ流动方式进行了改进，提出了第一壁环向流动（平行环向磁场流动），核燃料增殖区径向流动的ＭＨＤ流动的设计，以解决混合堆为改善堆的经济性而采取提高包层核燃料富集度的途径所速来的热工，ＭＨＤ压降和安全问题。分析和数值计算结果表明，第一壁环向流动设计可以满足包层核燃料富集度从０．５％增加到１％，相应的热功率从４５００ＭＷ增加到...     

Liquid lithium divertor system for fusion reactor

One of the most critical issues for the steady state fusion reactor is the heat flux in the divertor target. This paper proposes a liquid lithium divertor system to solve this problem. The proposed divertor system consists of a liquid lithium target, an evaporation chamber and a differential evacuation chamber. The heat coming from the fusion plasma along the divertor leg is removed by evaporation of lithium. The lithium vapor is condensed on the wall and is circulated with a pump. The coolant temperature for the wall is high enough to drive a power generator. Narrow slits along the divertor leg and the differential evacuation chamber reduce leakage of lithium vapor to the plasma chamber. A preliminary estimation predicts that the lithium ion density in the core plasma is lower than the plasma density.     

Maintenance Method for Tokamak Fusion Reactors with Downward Access to Torus Inboard Structures

A concept of tokamak fusion reactor maintenance is presented. Reactor structures and maintenance machines are arranged so that the component inside a shielding structure can be replaced through the hatches located on the upper side of the torus shielding structure. The plasma vacuum boundary is constituted by the inside wall of the shielding structure. The magnet vacuum chamber contains two toroidal magnets in a single room, so that strong support structures can be placed between these toroidal magnets. A merit of this reactor is that the inboard reactor structures are accessible with keeping the magnet cryogenic condition and without disassembling any major reactor components. The practicability of this method will depend on the time required to move the blanket segments in the toroidal direction and to weld pipes by remote handling. A number of ideas for reducing this time are presented.     

Tests of the cryogenic target for lithium and hydrogen isotopes extraction from the chamber of T-11M tokamak without its venting

T-11M lithium program is focused on a solution of technological issues of a steady-state tokamak with liquid lithium plasma facing components (PFC). Lithium, collected by the chamber wall of such tokamak is able to capture a considerable amount of tritium, which is unacceptable. In order to restrict the level of lithium deposited on the chamber wall and captured tritium it was suggested early to use a cryogenic target technique. Such target placed in the plasma of glow discharge (GDH, He or Ar) during the tokamak conditioning can play the role of collector of lithium and tritium atoms which were sputtered by GD bombardment of the wall. The collected lithium and tritium can be evacuated mechanically together with target from tokamak chamber through vacuum lock without venting. Cryogenic target, cooled by liquid nitrogen (LN), was installed in the T-11M and tested in different modes of wall conditioning and tokamak operations. The maximum speed of the lithium collection during GDH was 3.5 mg/h, that corresponds “to contamination” of wall by lithium during approximately 200 regular shots of T-11M which are equivalent to two-week regular operations. It was established that considerable part of lithium was collected in ionized state. On this basis it can be suggested the creation in tokamak chamber an equivalent ionic pump for extraction both lithium and tritium from chamber without venting during regular tokamak operation.     

Modeling hydrogen and helium entrapment in flowing liquid metal surfaces as plasma-facing components in fusion devices

In a fusion reactor, the ability to use liquids as plasma-facing components (PFCs) depends on their interaction with the plasma and the magnetic field. One important issue for the moving liquid is the ability to entrain particles that strike the PFC surface (helium and hydrogen isotopes) while accommodating high heat loads. To study this problem, an analytical model and a two-dimensional comprehensive numerical model have been developed and implemented in the HEIGHTS computer simulation package. The models take into account the kinetics of particle injection, motion and interactions with the liquid lattice, and the ultimate release from the surface. The models were used to investigate an important issue, whether He particles can be pumped by the PFC liquid rather than requiring a standard vacuum system. Hydrogen isotope (DT) particles that strike the surface will likely be trapped in the liquid-metal surface (e.g., lithium) due to the high chemical solubility of hydrogen. The impinging He particles in the established low-recycling regime at PFCs could be harder to pump using the standard vacuum pumping techniques. The analysis results indicate a reasonable chance of adequate helium self-trapping in flowing lithium as PFC without active pumping.     

Liquid metal magnetohydrodynamic flows in manifolds of dual coolant lead lithium blankets

An attractive blanket concept for a fusion reactor is the dual coolant lead lithium (DCLL) blanket where reduced activation steel is used as structural material and a lead lithium alloy serves both to produce tritium and to remove the heat in the breeder zone. Helium is employed to cool the first wall and the blanket structure.Some critical issues for the feasibility of this blanket concept are related to complex induced electric currents and 3D magnetohydrodynamic (MHD) phenomena that occur in distributing and collecting liquid metal manifolds. They can result in large pressure drop and undesirable flow imbalance in parallel poloidal ducts forming blanket modules.In the present paper liquid metal MHD flows are studied for different design options of a DCLL blanket manifold with the aim of identifying possible sources of flow imbalance and to predict velocity and pressure distributions.     

Analysis of the reference accidental sequence for safety assessment of LLCB TBM system

Safety analysis of the reference accidental sequence has been carried out for Lead Lithium cooled Ceramic Breeder (LLCB) Test Blanket Module (TBM) system; India's prototype of DEMO blanket concept for testing in International Thermonuclear Experimental Reactor (ITER). The accidental event analyzed starts with a Postulated Initiating Event (PIE) of ex-vessel loss of first wall helium coolant due to guillotine rupture of coolant pipe with simultaneous assumed failure of plasma shutdown system. Three different variants of the sequences analyzed include simultaneous additional failures of TBM and ITER first wall, failure of TBM box resulting in to spilling of lead lithium liquid metal in to vacuum vessel and reactor trip on Loss of Coolant Accident (LOCA) signal from TBM system. The analysis address specific reactor safety concerns, such as pressurization of confinement buildings, vacuum vessel pressurization, release of activated products and tritium during these accidental events and hydrogen production from chemical reactions between lead–lithium liquid metal and beryllium with water. An in-house customized computer code is developed and through these deterministic safety analyses the prescribed safety limits are shown to be well within limits for Indian LLCB-TBM design and it also meets overall safety goal for ITER. This paper reports transient analysis results of the safety assessment.     

Recent progress of NSTX lithium program and opportunities for magnetic fusion research

Lithium wall coating techniques have been experimentally explored on National Spherical Torus Experiment (NSTX) for the last six years. The lithium experimentation on NSTX started with a few milligrams of lithium injected into the plasma as pellets and it has evolved to a dual lithium evaporation system which can evaporate up to ～160 g of lithium onto the lower divertor plates between re-loadings. The unique feature of the NSTX lithium research program is that it can investigate the effects of lithium coated plasma-facing components in H-mode divertor plasmas. This lithium evaporation system has produced many intriguing and potentially important results. In 2010, the NSTX lithium program has focused on the effects of liquid lithium divertor (LLD) surfaces including the divertor heat load, deuterium pumping, impurity control, electron thermal confinement, H-mode pedestal physics, and enhanced plasma performance. To fill the LLD with lithium, 1300 g of lithium was evaporated into the NSTX vacuum vessel during the 2010 operations. The routine use of lithium in 2010 has significantly improved the plasma shot availability resulting in a record number of plasma shots in any given year. In this paper, as a follow-on paper from the 1st lithium symposium [1], we review the recent progress toward developing fundamental understanding of the NSTX lithium experimental observations as well as the opportunities and associated R&D required for use of lithium in future magnetic fusion facilities including ITER.     

3D thermo-fluid MHD simulation of single straight channel flow in LLCB TBM

India is developing lead lithium cooled ceramic breeder (LLCB) blanket for its DEMO fusion reactor. The mock-up blanket (TBM), using this concept, will be tested in ITER for its tritium breeding and high-grade heat extraction efficiency. In this TBM, pressurized helium is used to remove the heat from first wall, top and bottom plates of TBM. The Pb–Li is used to extract heat from the breeder zones. The flow of Pb–Li with average velocity 0.1 m/s inside the channel can be significantly modified due to MHD effects, which arise because of the presence of strong toroidal magnetic field. A numerical approach is established to capture this flow modification at higher Hartmann numbers (≥20,000). As a validation part of the developed code, MHD phenomenon is studied in 2-D square geometry and numerically obtained velocity profile is compared with available Hunt's analytical results. Thermo-fluid MHD analysis using this code, has been carried out for single rectangular duct of LLCB TBM. The heat transfer has been studied by keeping hot breeders at both sides of the flow channel. The results suggest modification in steady state MHD velocity profile as the liquid flows along the flow length. However, the temperature in various zone remains well within the maximum allowable limit.     

流动锂液帘对堆芯等离子体影响的评估(英文)

运用零维模型评估了流动液态锂幕帘作为聚变实验增殖堆工程概要设计 (FEB-E) 第一壁对聚变等离子体的影响。得到了锂液帘工作温度对堆芯有效平均等离子体电荷?Zeff?,燃料稀释以及聚变功率之间的关系。表明在正常工作情况下,液态锂的蒸发对?Zeff?的影响不是很严重,但对燃料稀释和聚变功率的影响却较为敏感。作为一个例子,对较高功率密度的反剪切位形聚变实验增殖堆FEB-E设计方案 II,计算了液帘的流速与它表面最大温升的关系,结果表明:即便0.5m/s的低速流动液帘第一壁, 蒸发对聚变等离子体的影响也甚微。     

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.     

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.     

聚变增殖堆液态金属Li自冷包层的热工水力分析

完成了托卡马克商用混合堆 TCB(Tokamak Commercial Breeder)Li 自冷包层设计的热工水力分析,讨论了热工水力设计中的一些关键问题。用两维有限元热传导程序 AYER 计算了 TCB 包层的温度分布,用液态金属 MHD(Magnetohydraudynamic)压降公式计算了包层的压降。同时,还分析了包层冷却剂丧失事故 LOCA 的瞬态热工过程。分析表明,正常工况下,包层结构材料最高温度,结构材料与冷却剂界面最高温度,以及包层总压降都满足堆设计要求。在 LOCA 工况下,如果停堆后1小时内包层中的燃料球能够借助重力卸出包层,第一壁和包层是安全的,并且不会受到损伤。     

Evidence of vapor shielding effect on heat flux loaded on flowing liquid lithium limiter in EAST

A lithium (Li) vapour layer was formed around a flowing liquid Li limiter to shield against the plasma incident power and reduce limiter heat flux in the EAST tokamak. The results revealed that after a plasma operation of a few seconds, the layer became clear, which indicated a strong Li emission with a decrease in the limiter surface temperature. This emission resulted in a dense vapour around the limiter, and Li ions moved along the magnetic field to form a green shielding layer on the limiter. The plasma heat flux loaded on the limiter, measured by the probe installed on the limiter, was approximately 52% lower than that detected by a fast-reciprocating probe at the same radial position without the limiter in EAST. Additionally, approximately 42% of the parallel heat flux was dissipated directly with the enhanced Li radiation in the discharge with the liquid metal infused trenches (LIMIT) limiter. This observation revealed that the Li vapour layer exhibited an excellent shielding effect to liquid Li on plasma heat flux, which is a possible benefit of liquid-plasma-facing components in future fusion devices.