DEMO divertor maintenance

This paper presents a DEMO divertor segmentation and maintenance scheme together with maintenance time estimations. As far as it is known, it is the first such study for DEMO, and the work was coordinated by EFDA. The approach of the study was that DEMO divertor and its handling shall be very similar to that of ITER, therefore ITER divertor segmentation and maintenance schemes were used as starting point. The maintenance scheme for both ITER and DEMO is the following: the divertor is segmented into cassettes. For maintenance 3 divertor ports are used through which all equipment and cassettes are inserted and removed. A toroidal rail based machine transports the element toroidally in front of the port, from where a radial tractor is used that move it between cask and VV. The main difference between DEMO and ITER is the increase in major radius from 6.2 to 7.5 m, and the cassette segmentation changed from 6.66° to 11.25°. These two factors increase the cassette weight from 12 to 25 tons. Therefore the cantilevered radial transportation of ITER cassettes was not adapted, but instead, the first design of a new, bottom supported radial machine is proposed. For the entire blanket maintenance 70 days were calculated instead of the 179 days in ITER. It was also concluded, that if in-vessel toroidal machine is used for blanket maintenance, there are no real possibilities for parallel blanket/divertor replacement.     

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.     

Current status of He-cooled divertor development for DEMO

A He-cooled divertor concept for DEMO is being investigated at the Forschungszentrum Karlsruhe within the framework of the EU power plant conceptual study. The design goal is to resist a heat flux of 10 MW/m² at least. The major R&D areas are design, analyses, fabrication technology, and experimental design verification. A modular design is preferred for thermal stress reduction. The HEMJ (He-cooled modular divertor with multiple-jet cooling) was chosen as reference concept. It employs small tiles made of tungsten, which are brazed to a thimble made of tungsten alloy W-1%La₂O₃. The W finger units are connected to the main structure of ODS Eurofer steel by means of a copper casting with mechanical interlock. The divertor modules are cooled by helium jets (10 MPa, 600 °C) impinging onto the heated inner surface of the thimble.In cooperation with the Efremov Institute a combined helium loop & electron beam facility (60 kW, 27 keV) was built in St. Petersburg, Russia, for experimental verification of the design. It enables mock-up testing at a nominal helium inlet temperature of 600 °C, an internal pressure of 10 MPa, and a pressure difference in the mock-up of up to 0.5 MPa. Technological studies were performed on manufacturing of the W finger mock-ups. Several high heat flux tests were successfully performed till now. Post-examination and characterisation of the mock-ups subjected to the high heat flux tests were performed in collaboration with Forschungszentrum Jülich. Altogether, the test results confirm the divertor performance required. The helium-cooled divertor concept was demonstrated to be feasible. The knowledge gained from these experiments and some aspects on the design improvement are discussed in this contribution.     

Manufacturing and testing of a HETS module for DEMO divertor

The development of a divertor concept for fusion power plants that is able to grant efficient recovery and conversion of the considerable fraction (～15%) of the total fusion thermal power incident is deemed to be an urgent task to meet in the EU Fast Track scenario. The He-cooled conceptual divertor design is one of the possible candidates. Helium cooling offers several advantages including chemical and neutronic inertness and the ability to operate at higher temperatures and lower pressures than those required for water cooling. The HETS (high-efficiency thermal shield) concept, initially developed by ENEA for water, has been adapted for use with He as coolant. This DEMO divertor concept is based on elements joined in series and protected by a hemispheric dome; it allows an increase of thermal exchange coefficient both for high speed of gas and for “jet impingement” effects of gas coming out from the internal side of hemispheric dome. It has been calculated to be capable of sustaining an incident heat flux of 10 MW/m² when operating at 10 MPa, an inlet He temperature of 600 °C, and an outlet temperature of 800 °C. The presented activity, performed in the frame of EFDA-TW5TRP-001 task, was focused on the manufacturing of a single HETS module and on its thermal–hydraulic testing. The materials used for the HETS module manufacturing were all DEMO-compatible: W for the armor material and the hemispherical-dome, DENSIMET for the exchanger body. The testing is performed by connecting the module to HEFUS3 He loop system that is a facility able to supply the He flow to the required testing conditions: 400 °C, 4–8 MPa and 20–40 g/s. The needed incident heat flux is obtained by RF inducting equipment coupled to an inductor coil installed just over the HETS module. A CFD analysis by ANSYS-CFX was performed in order to predict the thermal–mechanical behavior of the module and a final comparison with the experimental data is required to validate the CFD results. All parameters are monitored and recorded by data acquisition system.     

Potential and limits of water cooled divertor concepts based on monoblock design as possible candidates for a DEMO reactor

In this paper water-cooled divertor concepts based on tungsten monoblock design identified in previous studies as candidate for fusion power plant have been reviewed to assess their potential and limits as possible candidates for a DEMO concept deliverable in a short to medium term (“conservative baseline design”). The rationale and technology development assumptions that have led to their selection are revisited taking into account present factual information on reactor parameters, materials properties and manufacturing technologies.For that purpose, main parameters impacting the divertor design are identified and their relevance discussed. The state of the art knowledge on materials and relevant manufacturing techniques is reviewed. Particular attention is paid to material properties change after irradiation; phenomenon thresholds (if any) and possible operating ranges are identified (in terms of temperature and damage dose). The suitability of various proposed heat sink/structural and sacrificial layer materials, as proposed in the past, are re-assessed (e.g. with regard to the possibility of reducing peak heat flux and/or neutron radiation damages). As a result, potential and limits of various proposed concepts are highlighted, ranges in which they could operate (if any) defined and possible improvements are proposed.Identified missing point in materials database and/or manufacturing techniques knowledge that should be uppermost investigated in future R&D activities are reported.This work has been carried out in the frame of EFDA PPPT Work Programme activities.     

Non-destructive examination of the bonding interface in DEMO divertor fingers

Plasma facing components (PFCs) with tungsten (W) armor materials for DEMO divertor require a high heat flux removal capability (at least 10 MW/m² in steady-state conditions). The reference divertor PFC concept is a finger with a tungsten tile as a protection and sacrificial layer brazed to a thimble made of tungsten alloy W – 1% La₂O₃ (WL10). Defects may be located at the W thimble to W tile interface. As the number of fingers is considerable (>250,000), it is then a major issue to develop a reliable control procedure in order to control with a non-destructive examination the fabrication processes. The feasibility for detecting defect with infrared thermography SATIR test bed is presented. SATIR is based on the heat transient method and is used as an inspection tool in order to assess component heat transfer capability. SATIR tests were performed on fingers integrating or not the complex He cooling system (steel cartridge with jet holes). Millimeter size artificial defects were manufactured and their detectability was evaluated. Results of this study demonstrate that the SATIR method can be considered as a relevant non-destructive technique examination for the defect detection of DEMO divertor fingers.     

Nuclear heating and helium production at the inboard divertor rail for DEMO

Monte Carlo simulations were carried out for the DEMO model. Distributions of both the nuclear heating and the helium production in the area between the blanket and the divertor were calculated with the MCNP5 code for the reference case, when the DEMO geometry was not changed. Next a segment of the divertor and the lower part of the manifold were modified. Two new arrangements were studied. The simulations show that for one of the examined cases the helium production and the nuclear heating can be reduced roughly three or even four times in the investigated area. Besides the nuclear heating and the He production were estimated at the fastener (bolt head). The use of the modified divertor and a rail protecting the blanket is essential in the DEMO design.     

Design and cooling of the edge segments of the DEMO divertor target plates

In support of shadowing of the divertor target plate edges in toroidal direction against damage caused by the incident particles, the fingers at the boundary of the target plate should ideally form a flat surface. The reference cooling fingers are of hexagonal shape and when assembled together, their edge boundary cannot be flat. Therefore, the boundary segments need to be designed in a different way. Three possible designs are investigated: non-symmetric pentagonal fingers and two square-shaped fingers of different sizes, all cooled by the same type of concentric cartridge as in the reference design. Their heat transfer performance is analyzed from the point of view of maximum allowable temperature of the thimble structure. The computational fluid dynamics (CFD) analysis is performed to obtain the minimum mass flow rate of the coolant which is necessary to keep the structure's temperature below the permissible limit at an acceptable pressure loss.     

Preliminary concept design of the divertor remote handling system for DEMO power plant

This paper is based on the remote maintenance system project (WPRM) for the demonstration fusion power reactor (DEMO). Following ITER, DEMO aims to confirm the capability of generating several hundred of MW of net electricity by 2050. The main objective of these activities is to develop an efficient and reliable remote handling (RH) system for replacing the divertor cassettes.This paper presents the preliminary results of the concept design of the divertor RH system. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections of 4 m each, and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel.Two alternative design of the end effector to grip and manipulate the divertor cassette are also presented in this work. Both the concepts are hydraulically actuated, basing on the ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate.The main objective of this paper is to illustrate the feasibility of DEMO divertor remote maintenance operations.     

Thermo-mechanical analysis of a DEMO divertor cooling finger under the EFREMOV test conditions

A divertor component of the forthcoming DEMO fusion reactor should be able to withstand heat flux loads larger than 10 MW/m². Successful design should withstand high flux loads for a number of load cycles since initially the DEMO reactor is expected to operate in a non-steady-state mode. Computations for evaluating the structural response of the divertor published so far have, however, been based on the stationary approach. A combined computational fluid dynamics and structural model for evaluating the structural response of a divertor under the non-stationary load conditions is therefore developed in this work. Heat transfer coefficients between the helium and inner surface of the thimble are calculated first by solving the helium steady-state flow equations. Spatially distributed heat transfer coefficients are then used as a boundary condition in a non-stationary thermo-mechanical analysis of the divertor. This analysis is performed for a number of load cycles under different surface heat flux levels. The model is validated against the EFREMOV test experimental conditions, designed to be close to reactor operation conditions. Good agreement of the highest temperatures on the tile’s top surface with the experimental data is obtained. The results suggest that there are three critical regions in the design where damage could initialize: (a) the thimble’s inner surface with the highest thermal gradients, (b) the tile’s outer surface and (c) the filler layer of the brazed tile-thimble joint where the temperature is higher than permissible. Post-examination data of experimental specimen confirm these conclusions as cracks were observed in the above mentioned areas (a) and (b), while melting of the layer (c) was also observed.     

Design of a water cooled monoblock divertor for DEMO using Eurofer as structural material

The performed investigation focus on a monoblock type design for a water cooled DEMO divertor using Eurofer as structural material. In 2013, a study case of such a concept was presented. It was shown that basic concepts using Eurofer as structural material are limited to an incident heat flux of 8 MW m⁻². Since, the EFDA agency issued new specifications. In this study, the conceptual design is reassessed with regard to specifications. Then, steady state thermal analyses and thermo-mechanical elastic analyses have been performed to define an upgrade of the geometry taking into account new specifications, design criteria and the maximum heat flux requirement of 10 MW m⁻². An analysis of the influence of each adjustable geometrical parameter on thermo-mechanical design criteria was performed. As a consequence, geometrical parameters were set in order to fit to materials requirements. For defined hydraulic conditions taken in the most favourable configuration, the limit of this design is estimated to an incident heat flux of 10 MW m⁻². Margin to critical heat flux and rules against progressive deformation/ratcheting in structural material limit the design.     

Influence of multiple jet cooling on the heat transfer and thermal stresses in DEMO divertor cooling finger

The divertor concept for DEMO fusion reactor is based on modular design cooled by multiple impinging jets. Such divertor should be able to withstand a surface heat flux of at least 10 MW/m² at an acceptable pumping power. To reduce the thermal loads the plasma-facing side of the divertor is build up of numerous small cooling fingers. Each cooling finger is cooled by an array of jets blowing through the holes on the steel cartridge.The size, number and arrangement of jets on the cartridge influences the heat transfer and pressure drop characteristics of the divertor. Five different cartridge designs are analyzed in the paper. The most critical parameters, such as structure temperature, heat removal ability, pressure drop, cooling efficiency and thermal stress loadings in the cooling finger are predicted for each cartridge design. A combined computational fluid dynamics and structural model was used to perform the necessary numerical analyses. The results have shown that the cartridge design with the best heat transfer and pressure drop characteristics is not also the most favorable choice from the point of view of minimum stress peaks.     

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.     

DEMO水冷包层第一壁结构优化设计研究

针对聚变示范堆（DEMO）水冷包层,通过计算流体力学程序CFX和计算结构力学程序ANSYSWorkbench中的SIMULATION模块进行单向流固耦合分析。在对现有设计的DEMO水冷包层第一壁温度和应力数值模拟分析的基础上,改变了第一壁流道结构,着重研究了不同流道结构下的温度和应力分布,分析了几何结构对最高温度和最大应力的影响,提出第一壁结构的优化设计方案。数值模拟结果表明,优化设计方案能有效降低第一壁结构中的最高温度和最大应力。     

Neutronic design analyses for a dual-coolant blanket concept: Optimization for a fusion reactor DEMO

The generation of design specifications for a DEMO reactor, including breeding blanket (BB), vacuum vessel (VV) and magnetic field coils (MFC), requires a consistent neutronic optimization of structures between plasma and MFC. This work targets iteratively to generate these neutronic specifications for a Dual-Coolant He/Pb15.7Li breeding blanket design. The iteration process focuses on the optimization of allowable space between plasma scrapped-off-layer and VV in order to generate a MFC/VV/BB/plasma sustainable configuration with minimum global system volumes. Two VV designs have been considered: (1) a double-walled option with light-weight stiffeners and (2) a thick massive one. The optimization process also involves VV materials, looking to warrant radiation impact operational limits on the MFC. The resulting nuclear responses: peak nuclear heating in toroidal field (TF) coil, tritium breeding ratio (TBR), power amplification factor and helium production in the structural material are provided.     

Assessment of DEMO challenges in technology and physics

The challenges that DEMO designs encounter in both technology and physics are reviewed. It is shown that it is very important to respect the interlinks between these fields when developing designs for DEMO. Examples for areas where such interlinks put very strict requirements are the development of a steady state tokamak operation scenario and the question of power exhaust taking into account the boundary conditions set by materials questions. Concerning steady state operation, we find that demands on the physics scenario are so high that pulsed operation of a tokamak DEMO should seriously be considered in conservative DEMO designs. Alternatively, the device could foresee a large fraction of externally driven current which calls for optimization of both plasma CD efficiency as well as wall plug efficiency of the CD system. In the exhaust area, a realistic estimate of the admissable time averaged peak heat flux at the target is of the order of 5 MW/m², leading to strict requirements for the operational scenario, which has to rely on an unprecedented high level of radiation loss by impurity seeding and the facilitation of partial detachment. Thus, exhaust scenarios along these lines have to be developed which are compatible with the confinement needs and the H-L back transition power for DEMO. In both areas, we discuss possible risk mitigation strategies based on conceptually different approaches.     

Pre-conceptual design assessment of DEMO remote maintenance

EDFA, as part of the Power Plant Physics and Technology programme, has been working on the pre-conceptual design of a Demonstration Power Plant (DEMO). As part of this programme, a review of the remote maintenance strategy considered maintenance solutions compatible with expected environmental conditions, whilst showing potential for meeting the plant availability targets. A key finding was that, for practical purposes, the expected radiation levels prohibit the use of complex remote handling operations to replace the first wall. In 2012/2013, these remote maintenance activities were further extended, providing an insight into the requirements, constraints and challenges. In particular, the assessment of blanket and divertor maintenance, in light of the expected radiation conditions and availability, has elaborated the need for a very different approach from that of ITER. This activity has produced some very informative virtual reality simulations of the blanket segments and pipe removal that are exceptionally valuable in communicating the complexity and scale of the required operations. Through these simulations, estimates of the maintenance task durations have been possible demonstrating that a full replacement of the blankets within 6 months could be achieved. The design of the first wall, including the need to use sacrificial limiters must still be investigated. In support of the maintenance operations, a first indication of the requirements of an Active Maintenance Facility (AMF) has been elaborated.     

Overview of EU DEMO design and R&D activities

One important objective of the EU fusion roadmap Horizon 2020 is to lay the foundation of a Demonstration Fusion Power Reactor (DEMO) to follow ITER, with the capability of generating several 100 MW of net electricity to the grid and operating with a closed fuel-cycle by 2050. This is currently viewed by many of the nations engaged in the construction of ITER as the remaining crucial step towards the exploitation of fusion power. This paper outlines the DEMO design and R&D approach that is being adopted in Europe and presents some of the preliminary design options that are under evaluation as well as the most urgent R&D work that is expected to be launched in the near-future. The R&D on materials for a near-term DEMO is discussed in detail elsewhere.     

MAST upgrade closed pumped divertor design and analysis

The design of the Mega Amp Spherical Tokamak (MAST) Upgrade project provides an excellent opportunity for advancements in tokamak physics in support of ITER and DEMO as well as CTF. One of the main areas of contribution is the divertor. MAST Upgrade will be capable of operating with the outer leg of the divertor being either conventional or with the novel Super-X Divertor (SXD) configuration, in which additional shaping of the magnetic field takes the divertor leg out to large radius. A region of low poloidal field in the chamber of the divertor allows the outer leg to traverse a considerable distance toroidally, reaching a large major radius. Cryopumps outboard the divertor targets will provide density control. This paper describes the current design and layout of the closed pumped divertor geometry to meet the configurations envisaged. The assembly supports symmetrical plasmas with both the upper and lower divertor assemblies able to accommodate either the conventional or SXD.     

Efficiency and availability driven R&D issues for DEMO

DEMO is the main step foreseen after ITER to demonstrate the technological and commercial viability of a fusion power plant. DEMO R&D requirements are usually identified on the basis of the functions expected from each individual system. An approach based on the analysis of overall plant functional requirements sheds new light on R&D needs. The analysis presented here focuses on two overall functional requirements, efficiency and availability. The results of this analysis are presented here putting emphasis on systems not sufficiently considered up to now, e.g. the heating and current drive systems, while more commonly addressed systems such as tritium breeding blankets are not discussed in detail. It is also concluded that an overall functional analysis should be adopted very early in the DEMO conceptual design studies in order to provide a fully integrated approach, which is an absolute requirement to ensure that the ambitious goals of this device will be ultimately met.