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.     

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.     

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.     

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.     

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.     

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。     

Upgrade of an integrated Langmuir probe system on the closed divertor target plates in the HL-2A tokamak

A newly designed divertor Langmuir probe diagnostic system has been installed in a rare closed divertor of the HL-2A tokamak and steadily operated for the study of divertor physics involved edge-localized mode mitigation, detachment and redistribution of heat flux, etc. Two sets of probe arrays including 274 probe tips were placed at two ports (approximately 180° separated toroidally), and the spatial and temporal resolutions of this measurement system could reach 6 mm and 1 μs, respectively. A novel design of the ceramic isolation ring can ensure reliable electrical insulation property between the graphite tip and the copper substrate plate where plasma impurities and the dust are deposited into the gaps for a long experimental time. Meanwhile, the condition monitoring and mode conversion between single and triple probe of the probe system could be conveniently implemented via a remote-control station. The preliminary experimental result shows that the divertor Langmuir probe system is capable of measuring the high spatiotemporal parameters involved the plasma density, electron temperature, particle flux as well as heat flux during the ELMy H-mode discharges.     

Conceptual design and heat transfer performance of a flat-tile water-cooled divertor target

The divertor target components for the Chinese fusion engineering test reactor (CFETR) and the future experimental advanced superconducting tokamak (EAST) need to remove a heat flux of up to ～20 MW m-2.In view of such a high heat flux removal requirement,this study proposes a conceptual design for a flat-tile divertor target based on explosive welding and brazing technology.Rectangular water-cooled channels with a special thermal transfer structure (TTS)are designed in the heat sink to improve the flat-tile divertor target's heat transfer performance(HTP).The parametric design and optimization methods are applied to study the influence of the TTS variation parameters,including height (H),width (W*),thickness (T),and spacing (L),on the HTP.The research results show that the flat-tile divertor target's HTP is sensitive to the TTS parameter changes,and the sensitivity is T ＞ L ＞ W* ＞ H.The HTP first increases and then decreases with the increase of T,L,and W* and gradually increases with the increase of H.The optimal design parameters are as follows:H =5.5 mm,W* =25.8 mm,T =2.2 mm,and L =9.7 mm.The HTP of the optimized flat-tile divertor target at different flow speeds and tungsten tile thicknesses is studied using the numerical simulation method.A flat-tile divertor mock-up is developed according to the optimized parameters.In addition,high heat flux (HHF)tests are performed on an electron beam facility to further investigate the mock-up HTP.The numerical simulation calculation results show that the optimized flat-tile divertor target has great potential for handling the steady-state heat load of 20 MW m-2 under the tungsten tile thickness＜5 mm and the flow speed ≥7 m s-1.The heat transfer efficiency of the flat-tile divertor target with rectangular cooling channels improves by ～ 13％ and ～30％ compared to that of the flat-tile divertor target with circular cooling channels and the ITER-like monoblock,respectively.The HHF tests indicate that the flat-tile divertor mock-up can successfully withstand 1000 cycles of 20 MW m-2 of heat load without visible deformation,damage,and HTP degradation.The surface temperature of the flat-tile divertor mock-up at the 1000th cycle is only ～930 ℃.The fiat-tile divertor target's HTP is greatly improved by the parametric design and optimization method,and is better than the ITER-like monoblock and the fiat-tile mock-up for the WEST divertor.This conceptual design is currently being applied to the engineering design of the CFETR and EAST flat-tile divertors.     

A study on nuclear analysis of the divertor region of the CFETR

This paper presents the nuclear analysis performance of the Chinese Fusion Engineering Test Reactor(CFETR)divertor region using the MCNP-5 Monte Carlo N-particles code in a 3D geometry model.We assessed the nuclear responses of the divertor region component systems and evaluated their shielding capability,which can support the development strategy of the physical and engineering design of the CFETR.Model specification based on the latest CAD model of the CFETR divertor has been integrated into the CFETR MCNP reference model with a major/minor radius R=7.2 m/a=2.2 m in the 22.5° model,and a fusion-power range of around 1-1.5 GW.The nuclear heating and radiation damage of the divertor system are enhanced compared to that of the ITER and the earlier CFETR design.The initial nuclear responses of the toroidal field coil and vacuum vessel systems showed that the shielding of the current divertor design is not sufficient and optimization work has been carried out.We also carried out calculations and analysis using a hypothetical operating scenario of over 14 years.An excellent improvement in the nuclear performance has been obtained by the improved additional shielding block in the divertor region when referring to the ITER design limit,which can support the design of the future update of the divertor region systems of the CFETR.     

Dynamic thermal-hydraulic modelling of the EU DEMO HCPB breeding blanket cooling loops

A global, system-level thermal-hydraulic model of the EU DEMO tokamak fusion reactor is currently under development and implementation in a suitable software at Politecnico di Torino, including the relevant heat transfer and fluid dynamics phenomena, which affect the performance of the different cooling circuits and components and their integration in a consistent design. The model is based on an object-oriented approach using the Modelica language, which easily allows to preserve the high modularity required at this stage of the design. The first module of the global model will simulate the blanket cooling system and will be able to investigate different coolant options and different cooling schemes, to be adapted to the different blanket systems currently under development in the Breeding Blanket (BB) project. The paper presents the Helium-Cooled Pebble Bed (HCPB) module of the EU DEMO blanket cooling loops system model. The model is used to compare different schemes for the cooling of the different components of the HCPB BB, and to suggest improvements aimed at optimizing the pumping power required by the cooling system. The model is then used to analyse a pulsed scenario, characteristic of the EU DEMO operation.     

Experimental and numerical investigation of a full-scale helium-cooled divertor finger mock-up

A modular helium-cooled divertor design based on the multi-jet impingement concept (HEMJ) that is capable of accommodating a surface heat flux of at least 10 MW/m² has been developed at the Forschungszentrum Karlsruhe. Experimental investigations with a full-scale mock-up designed and built at the Georgia Institute of Technology, Atlanta were carried out in the helium loop HEBLO at Karlsruhe. Tests were run at heat loads of up to 2 MW/m² and flow conditions of 38 °C and 8 MPa so that the Reynolds number matches that for the actual divertor operating conditions (21,600). Comparison between the experimental data and results of simulations performed using the computational fluid dynamics code ANSYS CFX showed good agreement for the cooled surface temperature distribution, while the pressure loss was underestimated by about 20% by the code.     

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.     

Study of safety features and accident scenarios in a fusion DEMO reactor

After the Fukushima Dai-ichi nuclear accident, a need for assuring safety of fusion energy has grown in the Japanese (JA) fusion research community. DEMO safety research has been launched as a part of Broader Approach DEMO Design Activities (BA-DDA). This paper reports progress in the fusion DEMO safety research conducted under BA-DDA. Safety requirements and evaluation guidelines have been, first of all, established based on those established in the Japanese ITER site invitation activities. The radioactive source terms and energies that can mobilize such source terms have been assessed for a reference DEMO concept. This concept employs in-vessel components that are cooled by pressurized water and built of a low activation ferritic steel (F82H), contains solid pebble beds made of lithium-titanate (Li₂TiO₃) and beryllium–titanium (Be₁₂Ti) for tritium breeding and neutron multiplication, respectively. It is shown that unlike the energies expected in ITER, the enthalpy in the first wall/blanket cooling loops is large compared to the other energies expected in the reference DEMO concept. Reference accident event sequences in the reference DEMO in this study have been analyzed based on the Master Logic Diagram and Functional Failure Mode and Effect Analysis techniques. Accident events of particular concern in the DEMO have been selected based on the event sequence analysis and the hazard assessment.     

Design and installation of the closed helical divertor in LHD

Neutral particle behavior in the Large Helical Device heliotron has been investigated to conduct the effective particle control using the intrinsic helical divertor. It was revealed that the torus in-out asymmetry observed in the neutral pressure distribution depended on the divertor particle flux distribution, and thus, on the operational magnetic configuration. It was also revealed that the neutral pressure in the vacuum vessel in LHD was below 0.1 Pa. Degradation of the plasma confinement with increasing of the neutral pressure was observed, and that suggested the effective particle control is necessary for the sustaining of long discharge with high performance plasma and the further improvement of the confinement. The modification of the open helical divertor to the closed one was investigated for the particle control using helical divertor by using EIRENE code. Results of the calculation showed that proper rearrangement of divertor plates and additional components, such as dome structure make the neutral particles to be compressed well in the divertor region, and effective divertor pumping to be possible. Based on the simulation and experimental results, design of the closed helical divertor was completed and it will be partially installed in the Large Helical Device before the experimental campaign in 2010.     

Influence of the X-point location on edge plasma transport in the J-TEXT tokamak with a high-field-side single-null divertor

High-density experiments in the high-field-side mid-plane single-null divertor configuration have been performed for the first time on J-TEXT.The experiments show an increase in the highest central channel line-averaged density from 2.73 x 1019 m-3 to 6.49 x 1019 m-3,while the X-point moves away from the target by increasing the divertor coil current.The corresponding Greenwald fraction rises from 0.50 to 0.79.For the impurity transport,the density normalized radiation intensity(absolute extreme ultraviolet and soft x-ray)of the central channel density decreased significantly(＞50％)with an increase in the plasma density.To better understand the underlying physics mechanisms,the 3D edge Monte Carlo code coupled with EIRENE(EMC3-EIRENE)has been implemented for the first time on J-TEXT.The simulation results show good agreement with the experimental findings.As the X-point moves away from the target,the divertor power decay length drops and the scrape-off layer impurity screening effect is enhanced.     

High current and low q95 scenario studies for FAST in the view of ITER and DEMO

The Fusion Advanced Study Torus (FAST) has been proposed as a possible European satellite, in view of ITER and DEMO, in order to: (a) explore plasma wall interaction in reactor relevant conditions, (b) test tools and scenarios for safe and reliable tokamak operation up to the border of stability, and (c) address fusion plasmas with a significant population of fast particles. A new FAST scenario has been designed focusing on low-q operation, at plasma current I_P = 10 MA, toroidal field B_T = 8.5 T, with a q₉₅ ≈ 2.3 that would correspond to I_P ≈ 20 MA in ITER. The flat-top of the discharge can last a couple of seconds (i.e. half the diffusive resistive time and twice the energy confinement time), and is limited by the heating of the toroidal field coils. A preliminary evaluation of the end-of-pulse temperatures and of the electromagnetic forces acting on the central solenoid pack and poloidal field coils has been performed. Moreover, a VDE plasma disruption has been simulated and the maximum total vertical force applied on the vacuum vessel has been estimated.     

Analysis of an ex-vessel break in the ITER divertor cooling loop

In the present work the integrated ECART code, developed for severe accident analysis in LWRs, is applied on the analysis of a large ex-vessel break in the divertor cooling loop of the international thermonuclear experimental reactor (ITER). A comparison of the ECART results with those obtained by Studsvik Nuclear AB (S), utilizing the MELCOR code, was also performed in the general framework of the quality assurance program for the ITER accident analyses. This comparison gives a good agreement in the results, both for thermal-hydraulics and the environmental radioactive releases. Mainly these analyses, from the point of view of the ITER safety, confirm that the accidental overpressure inside the vacuum vessel and the Tokamak cooling water system (TWCS) Vault is always well below the design limits and that the radioactive releases are adequately confined below the ITER guidelines.     

Analysis, verification, and benchmarking of the transient thermal hydraulic ITERTHA code for the design of ITER divertor

An analytical study for the International Thermonuclear Experimental Reactor Thermal Hydraulic Analysis code (ITERTHA) is carried out for a copper divertor with a 5 mm tungsten tile. The influence of the incident heat flux, swirl-tape insertion in cooling channels as well as the coolant flow velocity on the divertor thermal response is analyzed and discussed. The ITERTHA code results are verified by the commercial finite element code, COSMOS. The heat transfer coefficients at the nodes located on the cooling channel-wall are determined outside COSMOS code by the same methodology used in ITERTHA. A good agreement is achieved under different incident heat fluxes. The ITERTHA code is also benchmarked against the thermal-hydraulic calculation of the outer divertor of the Fusion Ignition Research Experiment, FIRE for an incident heat flux of 20 MW/m² and coolant flow velocity of 10 m/s in a cooling channel of 8 mm diameter with swirl-tape inserts of 2 ratio and 1.5 mm thickness. The results show excellent agreement for both steady and transient states and prove the successful implementation of both the hydraulic and heated diameters of the swirl-tape channels in the used heat transfer correlations.     

A low wall-loading DEMO reactor design with high priority for early and reliable realization of a tokamak fusion reactor over the cost performance

Based on scientific databases adopted for designing ITER plasmas and on the advancement of fusion nuclear technology from the recent R&D program, a low wall-loading DEMO fusion reactor has been designed, where high priority has been given to the early and reliable realization of a tokamak fusion plasma over the cost performance. Since the major radius of this DEMO reactor is chosen to be 10 m, plasma ignition is achievable with a low fusion power of 0.8 GW and an operation period of 4–5 hours is available only with inductive current drive. The low ignition power makes it possible to adopt a first wall with an austenitic stainless steel, for which significant databases and operating experience exists, due to its use in the presence of neutron irradiation in fission reactors. In step with development of advanced materials, a step-wise increase of the fusion power seems to be feasible and realistic, because this DEMO reactor has the potential to produce a fusion power of 5 GW.