Design studies on three-dimensional issues for liquid blanket systems in helical reactor FFHR

A new design activity is under way for a helical type DEMO reactor FFHR-d1. The first stage of the activity involves the fundamental issues related to three-dimensional blanket design: (1) the minimum blanket space required for reactor parameter decisions, (2) the support method for the helical blanket system, and (3) the blanket module design. Investigations have been performed with neutronics and mechanical finite-element method calculations. Neutronics investigations indicate that a tungsten carbide radiation shield could reduce the minimum blanket space requirement by ～30 cm at the inboard region of FFHR-d1 compared with the blanket space of ～100 cm in the previous FFHR2 design. The investigations also showed that main shielding materials, ferritic steel and B₄C, could be used separately in a two-layered shielding configuration. The ferritic steel layer of the radiation shield is considered suitable to support the helical blanket system instead of relying on a thin vacuum vessel of the helical reactor. A size of a blanket module for a replacement process and the preferable cooling channel direction under a magnetic field are also discussed.     

Waste management strategy focused on maintenance,storage and recycling

Waste is generated at the moment when the operation of a fusion reactor is halted and maintenance is started for periodic replacement of blanket modules and divertor. Used blanket and divertor need to be replaced shortly after the shutdown for high plant availability, as long as high surface dose rate and decay heat of the blanket and divertor can be handled. In this sense, nuclear characteristics of the blanket and divertor need to be understood for a reasonable maintenance scheme. For the purpose, neutronic calculations were carried out on the blanket and divertor using a THIDA-2 code with FENDL-2.0. For a SlimCS DEMO reactor, the calculated decay heat for each 1/12-sector was as high as 5 MW just after the shutdown and 0.3 MW one month later. For the maintenance, a cooled shielding structure (CSS) was proposed to remove the decay heat and to shield gamma-rays from the sector. When maintenance is done one month after the shutdown, the sector temperature is maintained to be 550 °C or lower with the cooling by the CSS of 50 °C. In order to avoid tritium release from the sector during the maintenance, a cask should be used to transport the sector. For efficient use of resources, breeding and neutron multiplying materials should be reused or recycled. A possible strategy for reuse or recycle is also presented.     

Assessing the feasibility of a high-temperature,helium-cooled vacuum vessel and first wall for the Vulcan tokamak conceptual design

The Vulcan conceptual design (R = 1.2 m, a = 0.3 m, B₀ = 7 T), a compact, steady-state tokamak for plasma–material interaction (PMI) science, must incorporate a vacuum vessel capable of operating at 1000 K in order to replicate the temperature-dependent physical chemistry that will govern PMI in a reactor. In addition, the Vulcan divertor must be capable of handling steady-state heat fluxes up to 10 MW m^?2 so that integrated materials testing can be performed under reactor-relevant conditions. A conceptual design scoping study has been performed to assess the challenges involved in achieving such a configuration. The Vulcan vacuum system comprises an inner, primary vacuum vessel that is thermally and mechanically isolated from the outer, secondary vacuum vessel by a 10 cm vacuum gap. The thermal isolation minimizes heat conduction between the high-temperature helium-cooled primary vessel and the water-cooled secondary vessel. The mechanical isolation allows for thermal expansion and enables vertical removal of the primary vessel for maintenance or replacement. Access to the primary vessel for diagnostics, lower hybrid waveguides, and helium coolant is achieved through ～1 m long intra-vessel pipes to minimize temperature gradients and is shown to be commensurate with the available port space in Vulcan. The isolated primary vacuum vessel is shown to be mechanically feasible and robust to plasma disruptions with analytic calculations and finite element analyses. Heat removal in the first wall and divertor, coupled with the ability to perform in situ maintenance and replacement of divertor components for scientific purposes, is achieved by combining existing helium-cooled techniques with innovative mechanical attachments of plasma facing components, either in plate-type helium-cooled modules or independently bolted, helium-jet impingement-cooled tiles. The vacuum vessel and first wall design enables a wide range of potential PFC materials and configurations to be tested with relative ease, providing a new approach to reactor-relevant PMI science.     

Optimization and limitations of known DEMO divertor concepts

In this work we will introduce and discuss improvements for two types of DEMO divertors based on known designs: (i) gas cooled designs and (ii) liquid coolant concepts. In a first step, the advantages and disadvantages of gas cooling as well as the necessity of a jet impingement to increase the heat transfer coefficients will be discussed. Further discussion deals with the pros and cons of liquid coolant concepts, like for example, liquid metal or water cooling.Thereafter, we will present two rather contrary DEMO divertor concepts which are based on today's knowledge on refractory materials science, fabrication and joining technology. The first improved concept uses water flowing through steel pipes, typically made of Eurofer steel. It is well known that using Eurofer at low temperatures is critical due to its severe embrittlement under neutron irradiation. Here we make a proposal how it could be possible to use the Eurofer steel anyway: the solution could consist in a limited operation period followed by an annealing cycle at 550 °C for a few hours during any maintenance shut down phases. The second design is based on the known helium cooling concept using jet impingement. Drawbacks of the actual He-cooled divertor design are small scale parts as well as the necessary high helium inlet temperature of about 600–800 °C which leads to the question: How can we deal with such high helium temperatures? This paper shows a solution for large scale components as well as a new thermal management for the helium outlet gas that we call ‘cooling of the coolant’.Both concepts are discussed in terms of materials selection due to material limits and joining technology with a special focus on the material issue using already existing and available materials.     

2-D temperature distribution and heat flux of PFC in 2011 KSTAR campaign

KSTAR has reached a plasma current up to 630 kA, plasma duration up to 12 s, and has achieved high confinement mode (H-mode) in 2011 campaign. The heat flux of PFC tile was estimated from the temperature increase of PFC since 2010. The heat flux of PFC tiles increases significantly with higher plasma current and longer pulse duration. The time-averaged heat flux of shots in 2010 campaign (with 3 s pulse durations and I_p of 611 kA) is 0.01 MW/m² while that in 2011 campaign (with 12 s pulse duration and I_p of 630 kA) is about 0.02 MW/m². The heat flux at divertor is 1.4–2 times higher than that at inboard limiter or passive stabilizer. With the cryopump operation, the heat flux at the central divertor is higher than that without cryopump. The heat flux at divertor is proportional to, of course, the duration of H-mode. Furthermore, a software tool, which visualizes the 2D temperature distribution of PFC tile and estimates the heat flux in real time, is developed.     

Powder Injection Molding – An innovative manufacturing method for He-cooled DEMO divertor components

At Karlsruhe Institute of Technology (KIT), a He-cooled divertor design for future fusion power plants has been developed. This concept is based on the use of modular cooling fingers made from tungsten and tungsten alloy, which are presently considered the most promising divertor materials to withstand the specific heat load of 10 MW/m². Since a large number of the finger modules (n > 250,000) are needed for the whole reactor, developing a mass-oriented manufacturing method is indispensable. In this regard, an innovative manufacturing technology, Powder Injection Molding (PIM), has been adapted to W processing at KIT since a couple of years. This production method is deemed promising in view of large-scale production of tungsten parts with high near-net-shape precision, hence, offering an advantage of cost-saving process compared to conventional machining.The complete technological PIM process for tungsten materials and its application on manufacturing of real divertor components, including the design of a new PIM tool is outlined and, results of the examination of the finished product after heat-treatment are discussed. A binary tungsten powder feedstock with a solid load of 50 vol.% was developed and successfully tested in molding experiments. After design, simulation and manufacturing of a new PIM tool, real divertor parts are produced. After heat-treatment (pre-sintering and HIP) the successful finished samples showed a sintered density of approximately 99%, a hardness of 457 HV0.1, a grain size of approximately 5 μm and a microstructure without cracks and porosity.     

Tiles chamfering and power handling of the MK II HD divertor

The JET high triangularity (δ, HD) divertor is an upgrade of the present JET divertor consisting of two modified toroidal segments which are: a new load bearing septum replacement plate (LB-SRP) tile located in the center of the divertor and a high field gap closure (HFGC) tile protecting inboard diagnostic cabling. The aim of the upgrade is to allow high power operation and a wider range of plasma triangularities at the divertor poloidal null. This paper describes the optimisation of the tile chamfering (including edge shadowing) and the power handling evaluation for a set of 12 planned plasma configurations given by the JET team and on two sets of mechanical tile tolerances issued by the JET drawing office. The PROTEUS code (magnetic equilibrium by finite element) is used to calculate the various field line angles, which are inputs for the chamfering angle calculation process. After calculating the chamfering angle values of each face, a checking exercise has been realised on the 3D CATIA models of the tiles by putting them at their extreme tolerance positions and validating if the shadowing is ensured for a angle calculated to take into account the worst possibilities. With the final chamfering angle value for each face, the power handling of the tiles has been estimated with finite element calculations. Power handling is given either by the critical time to reach 1800 °C at the tile surface for a total injected power of 40 MW or by the maximum total injected power allowable for a 10 s power pulse without exceeding 1800 °C. The estimated power handling gives promising results in regard to the JET EP project objectives.     

Surface chemistry analysis of lithium conditioned NSTX graphite tiles correlated to plasma performance

Lithium wall conditioning in NSTX has resulted in reduced divertor recycling, improved energy confinement, and reduced frequency of edge-localized modes (ELMs), up to the point of complete ELM suppression. NSTX tiles were removed from the vessel following the 2008 campaign and subsequently analyzed using X-ray photoelectron spectroscopy as well as nuclear reaction ion beam analysis. In this paper we relate surface chemistry to deuterium retention/recycling, develop methods for cleaning of passivated NSTX tiles, and explore a method to effectively extract bound deuterium from lithiated graphite. Li–O–D and Li–C–D complexes characteristic of deuterium retention that form during NSTX operations are revealed by sputter cleaning and heating. Heating to ～850 °C desorbed all deuterium complexes observed in the O 1s and C 1s photoelectron energy ranges. Tile locations within approximately ±2.5 cm of the lower vertical/horizontal divertor corner appear to have unused LiO bonds that are not saturated with deuterium, whereas locations immediately outboard of this region indicate high deuterium recycling. X-ray photo electron spectra of a specific NSTX tile with wide ranging lithium coverage indicate that a minimum lithium dose, 100–500 nm equivalent thickness, is required for effective deuterium retention. This threshold is suspected to be highly sensitive to surface morphology. The present analysis may explain why plasma discharges in NSTX continue to benefit from lithium coating thickness beyond the divertor deuterium ion implantation depth, which is nominally <10 nm.     

Towards assembly completion and preparation of experimental campaigns of Wendelstein 7-X in the perspective of a path to a stellarator fusion power plant

The superconducting stellarator device Wendelstein 7-X, currently under construction, is the key device for the proof of stellarator optimization principles. To establish the optimized stellarator as a serious candidate for a fusion reactor, reactor-relevant dimensionless plasma parameters must be achieved in fully integrated steady-state scenarios. After more than 10 years of construction time, the completion of the device is now approaching rapidly (mid-2014). We discuss the most important lessons learned during the device assembly and first experiences with coming major work packages. Those are (a) assembly of about 2500 large, water-cooled, 3d-shaped in-vessel component elements; (b) assembly of in total 14 superconducting current leads, one pair for each coil type; and (c) assembly of the device periphery including diagnostics and heating systems. In the second part we report on the present status of planning for the first operation phase (5–10 s discharge duration at 8 MW heating power), the completion and hardening of the device for full power steady-state operation, and the second operation phase (up to 30 min discharge duration at 10 MW heating power). It is the ultimate goal of operation phase one to develop credible and robust discharge scenarios for the high-power steady-state operation phase two. Beyond the improved equilibrium, confinement, and stability properties owing to stellarator optimization, this requires density control, impurity control, edge iota control as well as high density microwave heating. Of paramount importance is the operation of the island divertor, which is realized in the first operation phase as an inertially cooled conventional graphite target divertor. It will be replaced later on by the steady-state capable island divertor with its water-cooled carbon fiber reinforced carbon target elements.     

CFETR第一壁及赤道面外包层中子辐照损伤初步分析

中国聚变工程实验堆(Chinese Fusion Engineering Testing Reactor,CFETR)的包层和偏滤器第一壁面向堆芯等离子体,第一壁辐照损伤分析对于托克马克安全运行至关重要。赤道面外包层较其它包层距离堆芯等离子体中心更近,其结构材料承受中子辐照大。因此,进行中子辐照损伤评估十分必要。基于此目的,采用计算机辅助设计(Computer Aided Design,CAD)模型和蒙特卡罗中子学建模转换接口Mc CAD完成中子学建模,并用蒙特卡罗方法的粒子输运程序计算第一壁和氦冷固态外包层结构材料辐照损伤。此外,对比了铍和钨作为面向等离子体材料两种情况下第一壁的受损情况。计算结果表明,氦冷固态包层模型下结构材料可以满足CFETR一期的运行要求。     

Active radiative liquid lithium divertor concept

Developing a reactor compatible divertor has been identified as a particularly challenging technology problem for magnetic confinement fusion. Application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising Li results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor (RLLD) concept [1]. In the RLLD, Li is evaporated from the liquid lithium (LL) coated divertor strike point surface due to the intense heat flux. The evaporated Li is readily ionized by the plasma due to its low ionization energy, and the poor Li particle confinement near the divertor plate enables ionized Li ions to radiate strongly, resulting in a significant reduction in the divertor heat flux. This radiative process has the desired effect of spreading the localized divertor heat load to the rest of the divertor chamber wall surfaces, facilitating divertor heat removal. The modeling results indicated that the Li radiation can be quite strong, so that only a small amount of Li (∼a few mol/s) is needed to significantly reduce the divertor peak heat flux for typical reactor parameters. In this paper, we examine an active version of the RLLD, which we term ARLLD, where LL is injected in the upstream region of divertor. We find that the ARLLD has similar effectiveness in reducing the divertor heat flux as the RLLD, again requiring only a few mol/s of LL to significantly reduce the divertor peak heat flux for a reactor. An advantage of the ARLLD is that one can inject LL proactively even in a feedback mode to insure the divertor peak heat flux remains below an acceptable level, providing the first line of defense against excessive divertor heat loads which could result in damage to divertor PFCs. Moreover, the low confinement property of the divertor (i.e., <1 ms for Li particle confinement time) makes the ARLLD response fast enough to mitigate the effects of possible transient events such as large ELMs.     

Design and concept validation of the new solid tungsten divertor for ASDEX Upgrade

Div-III, a divertor with solid tungsten target tiles for ASDEX Upgrade is designed and tested and will be installed in 2013. It is a further step in exploring tungsten as material for plasma facing components. It avoids the restrictions of tungsten coatings on graphite and realizes an operation range up to 50 MJ energy removing capability in the outer divertor. In addition, it allows physics investigation such as erosion and deuterium retention as well as effects of castellation and target tilting. The design of the target itself and the attachment was optimized with FE-analysis and was intensively high heat tested up to a double overload. Cyclic tests reveal that the target and the attachment can be operated with the design load of 50 MJ without any damage. Even a twofold overload results in local recrystallization and minor cracks but the targets did not fail during operation. The redesign of the divertor structure was used to increase the conductance between the cryo-pump and the divertor region. The impact of the changed pumping efficiency was investigated with SOLPS/Eirene modeling. The modeling results are an indication for an easier access to lower SOL densities as expected for a higher pumping efficiency in the main chamber.     

The KSTAR plasma facing components for 2010 operation

The KSTAR plasma facing components (PFCs) consist of inboard limiter, poloidal limiter, divertor, passive stabilizer and neutral beam armor. The main function of the PFCs is to define boundary of operating plasma and to protect the vacuum vessel and in-vessel components such as diagnostic components, in vessel control coil and several kinds of launchers for heating and current drive systems. The divertor is designed to enhance effective particle control to keep high quality plasma with various flexibilities in the shaping control for wide range of operational regime. The passive stabilizer that is made of CuCrZr alloy is designed to passively control the vertical position and MHD instabilities during operation as well as outer boundary of the plasma. Since fabrication has been started for all of the plasma facing components from middle of 2009, the inboard limiter, the divertor, and the passive stabilizer were successfully installed in the vacuum vessel, in turn. Moreover, one set of neutral beam armor and three strings of poloidal limiters were also installed according to the heating system that newly comes in 2010. All the PFCs tiles were baked to 200 °C and the PFC system showed no vacuum leakage and other mechanical troubles. In this paper, key features, fabrication, results of assembly, and baking of the KSTAR PFCs are summarized in detail.     

Manufacturing and testing of W/Cu mono-block small scale mock-up for EAST by HIP and HRP technologies

ITER-like W/Cu mono-block plasma-facing components (PFCs) will be used in vertical target regions of the experimental advanced superconducting tokamak (EAST) divertor. The first W/Cu mono-block small scale mock-up with five W mono-blocks has been manufactured successfully by technological combination of hot isostatic pressing (HIP) and hot radial pressing (HRP). The joining of a W mono-block and a pure copper interlayer was achieved by means of HIP technology and the bonding strength was over 150 MPa. The good bonding between the pure copper interlayer and a CuCrZr cooling tube was obtained by means of HRP technology. In order to understand deeply the process of HRP, the stress distribution of the mock-up during HRP process was simulated using ANSYS code. Ultrasonic Nondestructive Testing (NDT) of the W/Cu and Cu/CuCrZr interfaces was performed, showing that excellent bonding of the W/Cu and Cu/CuCrZr interfaces. The thermal cycle fatigue testing of the mock-up has been carried out by means of an e-beam device in Southwest Institute of Physics, Chengdu (SWIP) and the mock-up withstood 1000 cycles of heat loads up to 8.4 MW/m² with the cooling water of 2 m/s, 20 °C, 0.2 MPa.     

Issues from activation/dose rates calculation for full W ITER machine and increased fluence

The paper presents and discusses the results of the activation calculations for ITER to investigate the effect of a possible increase of the neutron fluence on the First Wall (FW) and of the use of tungsten instead of beryllium as inboard and outboard FW protective layer. The new analyses based on the recent ITER design have been performed using the most recent, reliable and validated nuclear data and codes. Three situations have been considered: (1) average FW neutron fluence of 1.0 MWa/m² with beryllium FW protective layers (PLs); (2) average FW neutron fluence of 0.5 MWa/m² with tungsten FW PLs and (3) average FW neutron fluence of 1.0 MWa/m² with tungsten FW PLs.The used approach considers the Scalenea-1 radiation transport sequence for obtaining the 175 groups neutron fluxes in all the materials/zones on the radial direction of the ITER equatorial midplane and the EASY2007 code package for the material activation analysis. For tungsten PLs, calculations have also been performed using a MCNP approach in a 1D geometry. In this way the effect of the multi-group treatment of cross-sections is compared versus a continuous energy treatment and the deriving self-shielding effect is determined in the case of W where many resonances are present.     

Fabrication of a full-size mock-up for inboard 10° section of ITER vacuum vessel thermal shield

A full-scale mock-up of VVTS inboard section was made in order to validate its manufacturing processes before manufacturing the vacuum vessel thermal shield (VVTS) for ITER tokamak. VVTS inboard 10° section consists of 20 mm shells on which cooling tubes are welded and flange joints that connect adjacent thermal shield sectors. The whole VVTS inboard is divided into two by bisectional flange joint located at the center. All the manufacturing processes except silver coating were tested and verified in the fabrication of mock-up. For the forming and the welding, pre-qualification tests were conducted to find proper process conditions. Shell thickness change was measured after bending, forming and buffing processes. Shell distortion was adjusted after the welding. Welding was validated by non-destructive examination. Bisectional flange joint was successfully assembled by inserting pins and tightening with bolt/nut. Bolt hole margin of 2 mm for sector flange was revealed to be sufficient by successful sector assembly of upper and lower parts of mock-up. Handling jig was found to be essential because the inboard section was flexible. Dimensional inspection of the fabricated mock-up was performed with a 3D laser scanner.     

Design and test program of a simplified divertor dummy coil structure for the WEST project

In order to fully validate actively cooled tungsten plasma facing components (industrial fabrication, operation with long plasma duration), the implementation of a tungsten axisymmetric divertor structure in the tokamak Tore-Supra is studied. With this major upgrade, so-called WEST (Tungsten Environment in Steady state Tokamak), Tore-Supra will be able to address the problematic of long plasma discharges with a metallic divertor target.To do so, it is planned to install two symmetric divertor coils inside the vacuum vessel. This assembly, called divertor structure, is made up of two stainless steel casings containing a copper winding pack cooled by a pressurized hot water circuit (up to 180 °C, 4 MPa) and is designed to perform steady state plasma operation (up to 1000 s).The divertor structure will be a complex assembly ring of 4 m diameter representing a total weight of around 20 tons. The technical challenge of this component will be the implementation of angular sectors inside the vacuum vessel environment (TIG welding of the coil casing, induction brazing and electrical insulation of the copper winding). Moreover, this complex assembly must sustain harsh environmental conditions in terms of ultra high vacuum conditions, electromagnetical loads and electrical isolation (13 kV ground voltage) under high temperature.In order to fully validate the assembly and the performance of this complex component, the production of a scale one dummy coil is in progress.The paper will illustrate, the technical developments performed in order to finalize the design for the call for tender for fabrication. The progress and the first results of the simplified dummy coils will be also addressed.     

Thermal effects and component cooling in Magnum-PSI

Magnum-PSI is a linear plasma generator, built at the FOM-Institute for Plasma Physics Rijnhuizen. Subject of study will be the interaction of plasma with a diversity of surface materials. The machine is designed to provide an environment with a steady state high-flux plasma (up to 10²⁴ H⁺ ions/m² s) in a 3 T magnetic field with an exposed surface of 80 cm² up to 10 MW/m². Magnum-PSI will provide new insights in the complex physics and chemistry that will occur in the divertor region of the future experimental fusion reactor ITER and reactors beyond ITER. The conditions at the surface of the sample can be varied over a wide range, such as plasma temperature, beam diameter, particle flux, inclination angle of the target, background pressure and magnetic field. An important subject of attention in the design of the machine was thermal effects originating in the excess heat and gas flow from the plasma source and radiation from the target.     

Concept design on RH maintenance of CFETR Tokamak reactor

CFETR which stands for Chinese Fusion Engineering Testing Reactor is a superconducting Tokamak device. The concept design on RH maintenance of CFETR has been done in the past year. It is known that, the RH maintenance is one of the most important parts for Tokamak reactor. The fusion power was designed as 50–200 MW and its duty cycle time (or burning time) was estimated as 30–50%. The center magnetic field strength on the TF magnet is 5.0 T, the maximum capacity of the volt seconds provided by center solenoid winding will be about 160 VS. The plasma current will be 10 MA and its major radius and minor radius is 5.7 m and 1.6 m respectively. All the components of CFETR which provide their basic functions must be maintained and inspected during the reactor lifetime. Thus, the remote handling (RH) maintenance system should be a key component, which must be detailedly designed during the concept design processing of CFETR, for the operation of reactor. The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. What is more, the technical problems encountered in the design process will also be discussed.     

Progress of He-cooled divertor development for DEMO

A He-cooled divertor concept for DEMO [1] has been developed at Karlsruhe Institute of Technology (KIT) since a couple of years with the goal of reaching a heat flux of 10 MW/m² anticipated for DEMO. The reference concept HEMJ (He-cooled modular divertor with multiple-jet cooling) is based on the use of small cooling fingers – each composed of a tungsten tile brazed to a tungsten alloy thimble – as well as on impingement jet cooling with helium at 10 MPa, 600 °C. The cooling fingers are connected to the main structure of ODS Eurofer steel by brazing in combination with a mechanical interlock. This paper reports progress to date of the design accompanying R&Ds, i.e. primarily the fabrication technology and HHF experiments. For the latter a combined helium loop and electron beam facility (200 kW, 40 keV) at Efremov Institute, St. Petersburg, Russia, has been used. This facility enables mock-up testing at a nominal helium inlet temperature of 600 °C, a pressure of 10 MPa, and a maximal pressure head of 0.5 MPa. HHF test results till now confirm well the divertor design performance. In the recent test series in early 2010 the first breakthrough was achieved when a mock-up has survived over 1000 cycles at 10 MW/m² unscathed.