Evaluation of heat transfer by sublimation for the application to the divertor heat sink for high fusion energy conversion

Thermal and structural responses of divertor target were evaluated by using finite element method. High heat flux simulating ELMs at the level of 100 MW/m² was assumed onto the tungsten armor, and surface temperature profile was obtained. When dynamic heat load over 100 MW/m² was applied, the maximum surface temperature exceeded 1300 °C, and it caused recrystallization of tungsten regardless of the heat transfer below it. The result was used to conduct dynamic heat load experiment on tungsten, and material behavior of tungsten was evaluated under dynamic heat load. This study also proposed new concept of divertor heat sink which can distribute high heat flux and transfers the heat to high temperature medium. It consists of tungsten armor, composite enhanced with high thermal conductivity fiber, and heat transport system applying phase transition. High heat flux simulating ELMs was also applied to target surface of the divertor, temperature gradient, thermal stress of tungsten and composite were evaluated. Based on the results of analysis, thermal structural requirement was considered.     

Actively cooled plasma facing components for long pulse high power operation

This paper reviews the development of heat removal technology for plasma facing components (PFCs) and focuses on water-cooled PFCs for near term, high power applications and the use of the tungsten (W), carbon (C), and beryllium (Be) as the preferred armor materials. There are also brief summaries of developments in helium-cooled PFCs and applications of free liquid surfaces. Water-cooled PFCs with C armor have been developed for Tore Supra, ITER, LHD and W7-X. W or Be armor is of interest for ITER and other devices. W-coatings on graphite have been tried, and mockups with “W brush” armor, developed in the USA for ITER and emulated elsewhere, have withstood thermal cycling at 25 MW/m². Reliably joining of the armor has been a significant challenge. A shorter version of this paper was published previously in the International Toki Conference on Plasma Physics and Controlled Nuclear Fusion (ITC-10), Toki, Japan, January 18–21, 2000.     

Initial damage processes for diamond film exposure to hydrogen plasma

Diamond is considered to be a possible alternative to other carbon based materials as a plasma facing material in nuclear fusion devices due to its high thermal conductivity and resistance to chemical erosion. In this work CVD diamond films were exposed to hydrogen plasma in the MAGnetized Plasma Interaction Experiment (MAGPIE): a linear plasma device at the Australian National University which simulates plasma conditions relevant to nuclear fusion. Various negative sample stage biases of magnitude less than 500 V were applied to control the energies of impinging ions. Characterisation results from SEM, Raman spectroscopy and ERDA are presented. No measureable quantity of hydrogen retention was observed, this is either due to no incorporation of hydrogen into the diamond structure or due to initial incorporation as a hydrocarbon followed by subsequent etching back into the plasma. A model is presented for the initial stages of diamond erosion in fusion relevant hydrogen plasma that involves chemical erosion of non-diamond material from the surface by hydrogen radicals and damage to the subsurface region from energetic hydrogen ions. These results show that the initial damage processes in this plasma regime are comparable to previous studies of the fundamental processes as reported for less extreme plasma such as in the development of diamond films.     

Design and analysis of the W7-X divertor scraper element

Thehigh heat-flux divertor of the Wendelstein 7-X large stellarator experiment consists of 10 divertor units which are designed to carry a steady-state heat flux of 10 MW/m². However, the edge elements of this divertor are limited to only 5 MW/m², and may be overloaded in certain plasma scenarios. It is proposed to reduce this heat by placing an additional “scraper element” in each of the ten divertor locations. It will be constructed using carbon fiber composite (CFC) monoblock technology. The design of the monoblocks and the path of the cooling tubes must be optimized in order to survive the significant steady-state heat loads, provide adequate coverage for the existing divertor, be located within sub-millimeter accuracy, and take into account the boundaries to other in vessel components, all at a minimum cost. Computational fluid dynamics modeling has been performed to examine the thermal transfer through the monoblock swirl tube channels for the design of the monoblock orientation. An iterative physics modeling and computer aided design process is being performed to optimize the placement of the scraper element within the severe spatial restrictions.     

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.     

A neutron poison tritium breeding controller applied to a water cooled fusion reactor model

The generation of tritium in sufficient quantities is an absolute requirement for a next step fusion device such as DEMO due to the scarcity of tritium sources. Although the production of sufficient quantities of tritium will be one of the main challenges for DEMO, within an energy economy featuring several fusion power plants the active control of tritium production may be required in order to manage surplus tritium inventories at power plant sites. The primary reason for controlling the tritium inventory in such an economy would therefore be to minimise the risk and storage costs associated with large quantities of surplus tritium. In order to ensure that enough tritium will be produced in a reactor which contains a solid tritium breeder, over the reactor's lifetime, the tritium breeding rate at the beginning of its lifetime is relatively high and reduces over time. This causes a large surplus tritium inventory to build up until approximately halfway through the lifetime of the blanket, when the inventory begins to decrease. This surplus tritium inventory could exceed several tens of kilograms of tritium, impacting on possible safety and licensing conditions that may exist.This paper describes a possible solution to the surplus tritium inventory problem that involves neutron poison injection into the coolant, which is managed with a tritium breeding controller. A simple PID controller and is used to manage the injection of the neutron absorbing compounds into the water coolant of a stratified blanket model, depending on the difference between the required tritium excess inventory and the measured tritium excess inventory. The compounds effectively reduce the amount of low energy neutrons available to react with lithium compounds, thus reducing the tritium breeding ratio. This controller reduces the amount of tritium being produced at the start of the reactor's lifetime and increases the rate of tritium production towards the end of its lifetime. Thus, a relatively stable tritium production level may be maintained, allowing the control system to minimize the stored tritium with obvious safety benefits. The FATI code (Fusion Activation and Transport Interface) will be used to perform the tritium breeding and controller calculations.     

Tritium retention to the first wall of JT-60U

For economic and safety reasons, tritium (T) accumulation on plasma facing wall (PFW) of fusion reactor is strictly limited. In this study, T inventory in the graphite tiles used at the first wall of JT-60U was measured by a full combustion method. It was found that T was only retained near plasma facing surfaces sides of the tiles and the amount of retained T increased from <10¹¹ to <10¹³ T atoms/cm² with increasing the exposed discharge period of the tiles. Integrating the T retention with the total surface area of the outer first wall tiles, the fraction of retained T in that area was estimated to be 13% of the total T production. It was confirmed that these retained T were part of the energetic T produced by DD reactions without being replaced by HH discharges. Based on the retained fraction, the annual amount of T retention in the outer first wall of a demo-size reactor was calculated to be 360 g/burn-year at maximum. Even the value would be much less in the reactor, the accumulation of this kind of inventory could have significant contribution to the total T inventory.     

Recent US activities on advanced He-cooled W-alloy divertor concepts for fusion power plants

Several advanced He-cooled W-alloy divertor concepts have been considered recently for power plant applications. They range in scale from a plate configuration with characteristic dimension of the order of 1 m, to the ARIES-CS T-tube configuration with characteristic dimension of the order of 10 cm, to the EU FZK finger concept with characteristic dimension of the order of 1.5 cm. The trend in moving to smaller-scale units is aimed at minimizing the thermal stress under a given heat load; however, this is done at the expense of increasing the number of units, with a corresponding impact on the reliability of the system. The possibility of optimizing the design by combining different configurations in an integrated design, based on the anticipated divertor heat flux profile, also has been proposed. Several heat transfer enhancement schemes have been considered in these designs, including slot jet, multi-hole jet, porous media and pin arrays. This paper summarizes recent US efforts in this area, including optimization and assessment of the different concepts under power plant conditions. Analytical and experimental studies of the concepts and cooling schemes are presented. Key issues are identified and discussed to help guide future R&D, including fabrication, joining, material behavior under the fusion environment and impact of design choice on reliability.     

Ultrasonic techniques for quality assessment of ITER Divertor plasma facing component

The divertor is one of the most challenging components of ITER machine. Its plasma facing components contain thousands of joints that should be assessed to demonstrate their integrity during the required lifetime. Ultrasonic (US) techniques have been developed to study the capability of defect detection and to control the quality and degradation of these interfaces after the manufacturing process. Three types of joints made of carbon fibre composite to copper alloy, tungsten to copper alloy, and copper-to-copper alloy with two types of configurations have been studied. More than 100 samples representing these configurations and containing implanted flaws of different sizes have been examined.US techniques developed are detailed and results of validation samples examination before and after high heat flux (HHF) tests are presented. The results show that for W monoblocks the US technique is able to detect, locate and size the degradations in the two sample joints; for CFC monoblocks, the US technique is also able to detect, locate and size the calibrated defects in the two joints before the HHF, however after the HHF test the technique is not able to reliably detect defects in the CFC/Cu joint; finally, for the W flat tiles the US technique is able to detect, locate and size the calibrated defects in the two joints before HHF test, nevertheless defect location and sizing are more difficult after the HHF test.     

Prospects of High Temperature Superconductors for fusion magnets and power applications

During the last few years, progress in the field of second-generation High Temperature Superconductors (HTS) was breathtaking. Industry has taken up production of long length coated REBCO conductors with reduced angular dependency on external magnetic field and excellent critical current density jc. Consequently these REBCO tapes are used more and more in power application.For fusion magnets, high current conductors in the kA range are needed to limit the voltage during fast discharge. Several designs for high current cables using High Temperature Superconductors have been proposed. With the REBCO tape performance at hand, the prospects of fusion magnets based on such high current cables are promising. An operation at 4.5 K offers a comfortable temperature margin, more mechanical stability and the possibility to reach even higher fields compared to existing solutions with Nb₃Sn which could be interesting with respect to DEMO.After a brief overview of HTS use in power application the paper will give an overview of possible use of HTS material for fusion application. Present high current HTS cable designs are reviewed and the potential using such concepts for future fusion magnets is discussed.     

Performance of electro-plated and joined components for divertor application

A general challenge in divertor development, independently of design type and cooling medium water or helium, is the reliable and adapted joining of components. Depending on the design variants, the characteristics of the joints will be focused on functional or structural behavior to guarantee e.g. good thermal conductivity and sufficient mechanical strength. All variants will have in common that tungsten is the plasma facing material. Thus, material combinations to be joined will range from Cu base over steel to tungsten. Especially tungsten shows lacks in adapted joining due to its metallurgical behavior ranging from immiscibility over bad wetting up to brittle intermetallic phase formation. Joining assisted by electro-chemical deposition of functional and filler layers showed that encouraging progress was achieved in wetting applying nickel interlayers. Nickel proved to be a good reference material but alternative elements (e.g. Pd, Fe) may be more attractive in fusion to manufacture suitable joints.Replacing of Ni as activator element by Pd for W/W or W/steel joints was achieved and joining with Cu-filler was successfully performed. Manufactured joints were characterized applying metallurgical testing and SEM/EDX analyses demonstrating the applicability of Pd activator. First shear tests showed that the joints exhibit mechanical stability sufficient for technical application.     

Development of a hydrogen permeation sensor for future tritium applications

Tritium monitoring in lithium–lead eutectic is of great importance for the performance of liquid blankets in fusion reactors. In addition, tritium measurements will be required in order to proof tritium self-sufficiency in liquid metal breeding systems. On-line hydrogen (isotopes) sensors must be design and tested in order to accomplish these goals.In this work, an experimental set up was designed in order to test the permeation hydrogen sensors at 500 °C. This experimental set-up allowed working with controlled environments (different hydrogen partial pressures) and the temperature was measured using a thermocouple connected to a temperature controller that regulated an electrical heater.In a first set of experiments, a hydrogen sensor was constructed using an α-iron capsule as an active hydrogen area. The sensor was mounted and tested in the experimental set up. In a second set of experiments the α-iron capsule was replaced by a welded thin palladium disk in order to minimize the death volume. The experiments performed using both membranes (α-iron and palladium) showed that the operation of the sensors in the equilibrium mode required at least several hours to reach the hydrogen equilibrium pressure.     

Simulation of damage to tokamaks plasma facing components during intense abnormal power deposition

Intense power deposition on plasma facing components (PFC) is expected in tokamaks during loss of confinement events such as disruptions, vertical displacement events (VDE), runaway electrons (RE), or during normal operating conditions such as edge-localized modes (ELM). These highly energetic events are damaging enough to hinder long term operation and may not be easily mitigated without loss of structural or functional performance of the PFC. Surface erosion, melted/ablated-vaporized material splashing, and material transport into the bulk plasma are reliability-threatening for the machine and system performance.A novel particle-in-cell (PIC) technique has been developed and integrated into the existing HEIGHTS package in order to obtain a global view of the plasma evolution upon energy impingement. This newly developed PIC technique is benchmarked against plasma gun experimental data, the original HEIGHTS computer package, and laser experiments. Benchmarking results are shown in this paper for various relevant reactor and experimental devices. The evolution of the plasma vapor cloud is followed temporally and results are explained and commented as a function of the computational time needed and the accuracy of the calculation.     

Progress in development and application of lithium based components for Tokamak

Experiments with lithium plasma facing components (PFCs) show promising results for the operation of hot plasma facilities and the general improvement of plasma parameters. The design and development of new tokamak plasma facing material (PFM) based on lithium capillary porous systems (CPS) are described in this paper.The recent progress in the development of limiters with different kinds of CPS is relevant for protection of tokamak PFCs from damage under normal operation, ELMs and disruptions. New PFM eliminates the lithium flux into plasma, its pollution and lithium accumulation.Here we present an overview of the design and the experimental tests of the liquid lithium limiters. These limiters are based on CPS with hard matrix from stainless steel mesh, molybdenum and tungsten. Different types of limiter have been taken into account: the horizontal and vertical rail type limiters with passive and active cooling for investigation the possibility to provide the closed lithium circulation in tokamak chamber; the ring CPS-based limiter for investigation of lithium behavior in limiter scrape-off layer (SOL).Here we also present the preliminary results of the application of the cryogenic techniques for lithium removal from the chamber wall after operation in hot plasma.     

“PROCESS”: A systems code for fusion power plants—Part 1: Physics

PROCESS is a reactor systems code – it assesses the engineering and economic viability of a hypothetical fusion power station using simple models of all parts of a reactor system, from the basic plasma physics to the generation of electricity. It has been used for many years, but details of its operation have not been previously published. This paper describes some of its capabilities. PROCESS is usually used in optimisation mode, in which it finds a set of parameters that maximise (or minimise) a figure of merit chosen by the user, while being consistent with the inputs and the specified constraints. Because the user can apply all the physically relevant constraints, while allowing a large number of parameters to vary, it is in principle only necessary to run the code once to produce a self-consistent, physically plausible reactor model. The scope of PROCESS is very wide and goes well beyond reactor physics, including conversion of heat to electricity, buildings, and costs, but this paper describes only the plasma physics and magnetic field calculations.The capabilities of PROCESS in plasma physics are limited, as its main aim is to combine engineering, physics and economics. A model is described which shows the main plasma features of an inductive ITER scenario. Significant differences between the PROCESS results and the published scenario include the bootstrap current and loop voltage. The PROCESS models for these are being revised. Two new models for DEMO have been obtained. The first, DEMO A, is intended to be “conservative” in that it might be possible to build it using the technology of the near future. For example, since current drive technologies are not yet mature, only 12% of the current is assumed to be due to current drive. Consequently it is a pulsed machine, able to burn for only 1.65 hours at a time. Despite the comparatively large size (major radius is 9 m), the fusion power is only 1.95 GW. The assumed gross thermal efficiency is 33%, giving just 465 MW net electric power. The second, DEMO B, is intended to be “advanced” in that more optimistic assumptions are made. Comparison of DEMO A and B with a reference ITER scenario shows that current drive and bootstrap fraction need the most extrapolation from the perspective of plasma physics.     

Behaviour of Ti-doped CFCs under thermal fatigue tests

In spite of the remarkable progress in the design of in-vessel components for the divertor of the first International Thermonuclear Experimental Reactor (ITER), a great effort is still put into the development of manufacturing technologies for carbon armour with improved properties. Newly developed 3D titanium-doped carbon fibre reinforced composites and their corresponding undoped counterparts were brazed to a CuCrZr heat sink to produce actively cooled flat tile mock-ups. By exposing the mock-ups to thermal fatigue tests in an electron beam test facility, the material behaviour and the brazing between the individual constituents in the mock-up was qualified. The mock-ups with titanium-doped CFCs exhibited a significantly improved thermal fatigue resistance compared with those undoped materials. The comparison of these mock-ups with those produced using pristine NB31, one of the reference materials as plasma facing material for ITER, showed almost identical results, indicating the high potential of Ti-doped CFCs due to their improved thermal shock resistance.     

产氚回路中氚的在线监测研究

描述一套产氚回路中氚的在线监测装置——流气式正比计数系统。测试该系统的坪曲线、本底、探测效率等性能后,开展了氚在线监测实验研究。结果表明,该系统性能稳定,控制好测量条件,可作产氚回路中氚的在线监测设备。     

A method to produce lithium pellets for fueling and ELM pacing in NSTX-U

A device for producing small, high frequency spherical droplets or pellets for lithium or other liquid metals has been developed and could aid in the controlled excitation or pacing of edge-localized plasma modes (ELMs). The Liquid Lithium/metal Pellet Injector (LLPI) could also be used to replenish lithium coatings of plasma-facing components (PFCs) during a plasma discharge. With NSTX-U having longer pulse lengths (up to 5 s), it is desirable to be able to inject lithium during the discharge to maintain the beneficial effects. Using a nozzle injector design and a surrogate to lithium, Wood's metal, the LLPI has achieved droplet diameters between 0.6 mm < d_drop < 1 mm in diameter and frequencies up to 1.5 kHz with argon gas driving the formation. This paper presents the LLPI being developed with initial results mainly using Wood's metal and some lithium, using high pressure argon to force the liquid lithium through the nozzle.     

Theoretical study on fusion dynamics and evaporation residue cross sections for superheavy elements

The nuclear dynamical deformation, the fusion probability and the evaporation residue （ER） cross sections for the synthesis of superheavy nuclei are studied with the di-nuclear system model and the related dynamical potential energy surface. The intrinsic energy and the maximum dynamical deformations for 48Ca＋248Cm are calculated. The effect of dynamical deformation on the potential energy surface and fusion is investigated. It is found that the dynamical deformation influences the potential energy surface and fusion probability significantly. The dependence of the fusion probability on the angular momentum is investigated. The ER cross sections for some superheavy nuclei in 48Ca induced reactions are calculated and it is found that the theoretical results are in good agreement with the experimental results.     

Experimental study on the adsorption and desorption of tritium in the graphite materials for HTR-PM

Tritium behavior in the reactor such as production, diffusion and release are accompanied by their adsorption and desorption in graphite materials, which are essential to the safety of high temperature gas cooled reactor (HTGR). In order to study this important issue, hydrogen instead of tritium is experimentally used in this work and justified viable by theory. By performing multiple sets of comparative experiments, the features of hydrogen adsorption and desorption behavior changing by adsorption temperature and time in typical graphites used in HTR-PM (High Temperature Gas Cooled Reactor – Pebble Bed Module), i.e. reflective layer, fuel element and boron carbon bricks, have been observed and analyzed. Furthermore, the adsorption rates of hydrogen in the three materials as above at different conditions are also given. Based on the experimental results, tritium behavior in the HTR-PM was inferred and estimated, which is significant for the further study on the mechanism of tritium transport.