First Wall Design Research Articles

Design update and thermal-hydraulics of LLCB TBM first wall

The First Wall (FW) of Indian Lead Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) is a U-shaped structure made of India specific Reduced Activation Ferritic Martensitic Steel (IN-RAFMS). As the FW is directly exposed to high heat flux from plasma, it is actively cooled by high pressure and high temperature helium gas. Following the Conceptual Design Review (CDR) at ITER, some improvement on the design of LLCB TBM has been envisaged, this includes the FW design to reduce the operational load on helium cooling process systems. The optimization studies have been carried out selecting suitable channel size, shape and no. of flow circuits of FW to keep the FW temperature and pressure drop within the acceptable limits. The parametric study has been performed with different mass flow rates and channel configurations of FW. This analysis resulted in the reduction of mass flow requirement by about 38% of the current value. The optimized parameters have been used to carry out the thermal-hydraulic and thermo-mechanical analyses of the two selected configurations of FW using ANSYS code. The temperatures and stresses are found to be within the required limits. Finally, the detailed thermal-hydraulics of final selected design of TBM FW have been performed using ANSYS CFX to estimate temperature distribution along with the validation of analytically calculated pressure drop and Heat Transfer Coefficient (HTC). It is observed that the temperature distribution on FW obtained from the flow analysis in CFX and that obtained by performing thermal analysis in ANSYS using HTC from engineering correlations are similar. The detailed analysis of the variation of HTC along the flow length of the heat flux zone has also been discussed. This paper also discusses transient and accident thermal analysis of the updated design for some of the ITER scenarios and the temperatures have been found to be within the required limits.

Design, analysis and manufacturing of electrical strap for enhanced first wall panel

The JSC NIKIET is responsible for the manufacture of the ITER Enhanced Heat Flux (EHF) First Wall (FW) beam, the fingers bodies, the mechanical attachment system and electrical connection system of the FW panel to the shield block (SB). The Electrical strap (ES) is located on the FW rear surface and used for providing current through the FW to the SB and further to the Vacuum Vessel with a goal to reduce the current loops and consequently to decrease the electromagnetic loads acting on the FW during plasma disruption. The NIKIET specialists developed the new design of the electrical strap which has been implemented into the FW design. Electrical discharge machine (EDM) cutting is used for performing the 1.2mm thickness lamellas in the solid blank. This ES design has been approved by IO. The transient thermal analysis results show that temperature of ES is lower than 300°С for Inductive I operation mode and does not exceed allowable one. The fatigue strength criteria are satisfied for all the considered critical points of ES lamellas. Also the technological validation of ES manufacturing process has been successfully completed and full scale ES mock ups have been manufactured. This paper presents the design description of the EHF FW ES and FW bimetallic pedestal, results of thermal and structural analyses and first experience of mock up manufacturing that were performed by JSC NIKIET specialists in 2015.

New model of plasma heat load on the first wall

Scrape Off Layer (SOL) plasma circumnavigates the torus along magnetic field lines, and some part of it encounters a leading edge of the first wall (FW). This leads the concentration of the plasma energy in a small area, causing a hot spot. In a DEMO reactor that deals with a few hundreds of MW of plasma energy in the SOL, the leading edge problem is a critical issue in the FW design. For the design purposes in DEMO and power reactors, a new analysis method is introduced, where a flux tube is poloidally divided into the Apple Peel Like Elements (APPLEs) and each APPLE is enclosed by adjacent 4 magnetic field lines. Considering the contact area of the APPLEs on the FW and the radial energy transport between the APPLEs, FW heat load profile can be analyzed. The result of the case that the fusion power is 1.5GW, and the major radius is 8.2m, shows that the heat load is peaked to about 1MW/m2 near the inboard midplane and the baffle plate. This method is expected to be useful for the DEMO design.

Analysis results of the enhanced heat flux first wall panel elements

In the framework of Procurement Arrangement (PA) realization, specialists from NIKIET and Efremov Institute are re-assessing and optimizing the design of the First Wall (FW) Full Scale Prototype (FSP) in order to demonstrate its manufacturability and qualify critical technological processes. The design of the FW FSP is based on the FW 14 type A. The semi-prototype has been manufactured in order to verify the FW design. Based upon the manufacturing experience, some modifications have been implemented into the FSP design in order to simplify assembly and manufacturing processes.Each FW includes plasma-facing components (FW fingers), FW beam, mechanical attachment system and electrical connection system with the shield block. All the FW fingers are attached to the FW beam by two poloidal straight welds. In order to provide access of welding equipment, the 60mm slot is performed on the FW front surface. The central slot insert (CSI) is implemented into the FW design in order to protect the central slot from plasma radiation. The fixation of the CSI is provided by three circular welds and the hydraulic connection is provided by two pipes. The basic design of the CSI has been modified in order to simplify manufacturing and final assembly.This paper presents results of thermal and structural analysis of the modified FW design in particular CSI.

Numerical analysis of heat transfer in the first wall of CFETR WCSB blanket

China Fusion Engineering Test Reactor (CFETR), the first fusion reactor experiment project planned in China, is now being investigated in detail. Recently, a conceptual structural design of the Water-Cooled-Solid-Breeder (WCSB) blanket was proposed as one of the breeding blanket candidates for CFETR. In this research, based on the present design of the CFETR WCSB blanket, the heat transfer performance in the first wall (FW) under the pressurized water cooling condition was analyzed. The 3D computational fluid dynamics (CFD) results show that the maximal temperature of the FW will not exceed the limited temperature under normal or even higher heat flux condition. In addition, the effect of buoyancy on heat transfer is negligible under both conditions. The influence of roughness becomes increasingly important when the roughness height lies in the fully turbulent regime. The maximal temperature increases approximately linearly as the thickness of the front wall increases. It is also found that the heat flux and the local heat transfer coefficient are extremely non-uniform in the circumferential direction. Two 1D CFD methods are also evaluated by 3D CFD data, with the conclusion that both 1D results have some differences with the 3D data. The improved 1D method is more accurate than the former one. However, we ascertain that 1D methods should be used with caution for the water-cooling FW design.

Progress in the engineering design and assessment of the European DEMO first wall and divertor plasma facing components

The European DEMO power reactor is currently under conceptual design within the EUROfusion Consortium. One of the most critical activities is the engineering of the plasma-facing components (PFCs) covering the plasma chamber wall, which must operate reliably in an extreme environment of neutron irradiation and surface heat and particle flux, while also allowing sufficient neutron transmission to the tritium breeding blankets. A systems approach using advanced numerical analysis is vital to realising viable solutions for these first wall and divertor PFCs. Here, we present the system requirements and describe bespoke thermo-mechanical and thermo-hydraulic assessment procedures which have been used as tools for design. The current first wall and divertor designs are overviewed along with supporting analyses. The PFC solutions employed will necessarily vary around the wall, depending on local conditions, and must be designed in an integrated manner by analysis and physical testing.

Thermal-hydraulics of helium cooled First Wall channels and scoping investigations on performance improvement by application of ribs and mixing devices

The first wall (FW) of DEMO is a component with high thermal loads. The cooling of the FW has to comply with the material's upper and lower temperature limits and requirements from stress assessment, like low temperature gradients. Also, the cooling has to be integrated into the balance-of-plant, in a sense to deliver exergy to the power cycle and require a limited pumping power for coolant circulation. This paper deals with the basics of FW cooling and proposes optimization approaches. The effectiveness of several heat transfer enhancement techniques is investigated for the use in helium cooled FW designs for DEMO. Among these are wall-mounted ribs, large scale mixing devices and modified hydraulic diameter. Their performance is assessed by computational fluid dynamics (CFD), and heat transfer coefficients and pressure drop are compared. Based on the results, an extrapolation to high heat fluxes is tried to estimate the higher limits of cooling capabilities.

Conceptual design of divertor and first wall for DEMO-FNS

Key issues of design of the divertor and the first wall of DEMO-FNS are presented. A double null closed magnetic configuration was chosen with long external legs and V-shaped corners. The divertor employs a cassette design similar to that of ITER. Water-cooled first wall of the tokamak is made of Be tiles and CuCrZr-stainless steel shells. Lithium injection and circulation technologies are foreseen for protection of plasma facing components. Simulations of thermal loads onto the first wall and divertor plates suggest a possibility to distribute heat loads making them less than 10 MW m−2. Evaluations of sputtering and evaporation of plasma-facing materials suggest that lithium may protect the first wall. To prevent Be erosion at the outer divertor plates either the full detached divertor operation or arrangement of the renewal lithium flow on targets should be implemented. Test bed experiments on the Tsefey-M facility with the first wall mockup coated by Ве tiles and cooled by water are presented. The temperature of the surface of tiles reached 280–300 °С at 5 MW m−2 and 600–650 °С at 10.5 MW m−2. The mockup successfully withstood 1000 cycles with the lower thermal loading and 100 cycles with higher thermal loading.

Multiphysics Engineering Analysis for an Integrated Design of ITER Diagnostic First Wall and Diagnostic Shield Module Design

ITER diagnostic first walls (DFWs) and diagnostic shield modules (DSMs) inside the port plugs (PPs) are designed to protect diagnostic instrument and components from a harsh plasma environment and provide structural support while allowing for diagnostic access to the plasma. The design of DFWs and DSMs are driven by 1) plasma radiation and nuclear heating during normal operation 2) electromagnetic loads during plasma events and associate component structural responses. A multi-physics engineering analysis protocol for the design has been established at Princeton Plasma Physics Laboratory and it was used for the design of ITER DFWs and DSMs. The analyses were performed to address challenging design issues based on resultant stresses and deflections of the DFW-DSM-PP assembly for the main load cases. ITER Structural Design Criteria for In-Vessel Components (SDC-IC) required for design by analysis and three major issues driving the mechanical design of ITER DFWs are discussed. The general guidelines for the DSM design have been established as a result of design parametric studies.

Plasma facing components: a conceptual design strategy for the first wall in FAST tokamak

Satellite tokamaks are conceived with the main purpose of developing new or alternative ITER- and DEMO-relevant technologies, able to contribute in resolving the pending issues about plasma operation. In particular, a high criticality needs to be associated to the design of plasma facing components, i.e. first wall (FW) and divertor, due to physical, topological and thermo-structural reasons. In such a context, the design of the FW in FAST fusion plant, whose operational range is close to ITER’s one, takes place. According to the mission of experimental satellites, the FW design strategy, which is presented in this paper relies on a series of innovative design choices and proposals with a particular attention to the typical key points of plasma facing components design. Such an approach, taking into account a series of involved physical constraints and functional requirements to be fulfilled, marks a clear borderline with the FW solution adopted in ITER, in terms of basic ideas, manufacturing aspects, remote maintenance procedure, manifolds management, cooling cycle and support system configuration.

Preliminary electromagnetic, thermal and mechanical design for first wall and vacuum vessel of FAST

The fusion advanced study torus (FAST), with its compact design, high toroidal field and plasma current, faces many of the problems met by ITER, and at the same time anticipates much of the DEMO relevant physics and technology. The conceptual design of the first wall (FW) and the vacuum vessel (VV) has been defined on the basis of FAST operative conditions and of “Snow Flakes” (SF) magnetic topology, which is also relevant for DEMO.The EM loads are one of the most critical load components for the FW and the VV during plasma disruptions and a first dimensioning of these components for such loads is mandatory.During this first phase of R&D activities the conceptual design of the FW and VV have been assessed estimating, by means of FE simulations, the EM loads due to a typical vertical disruption event (VDE) in FAST. EM loads were then transferred on a FE mechanical model of the FAST structures and the mechanical response of the FW and VV design for the analyzed VDE event was assessed.The results indicate that design criteria are not fully satisfied by the current drawing of the VV and FW components.The most critical regions have been individuated and the effect of some geometrical and material changes has been checked in order to improve the structure.

Optimization of the first wall for the DEMO water cooled lithium lead blanket

The maximum heat load capacity of a DEMO First Wall (FW) of reasonable cost may impact the decision of the implementation of limiters in DEMO. An estimate of the engineering limit of the FW heat load capacity is an essential input for this decision. This paper describes the work performed to optimize the FW of the Water Cooled Lithium-Lead (WCLL) blanket concept for DEMO fusion reactor in order to increase its maximum heat load capacity.The optimization is based on the use of water at typical Pressurised Water Reactors conditions as coolant. The present WCLL FW with a waved plasma-faced surface and with circular channels was studied and the heat load limit has been predicted with FEM analysis equal to 1.0MWm−2 with respect to the Eurofer temperature limit.An optimization study was then carried out for a flat FW design considering thermal and mechanical constraints assuming inlet and outlet temperatures equal to 285°C/325°C respectively and based on geometric design parameters such as channel pitch, diameter of pipes and thicknesses. It became clear through the optimization that the advantages of a waved FW are diminished. Given the manufacturing issues of that concept, the waved FW was therefore not pursued further. Even if the optimization study shows that the maximum heat load could in principle be as high as 2.53MWm−2, it is reduced to 1.57MWm−2 when additional constraints are introduced in order not to affect corrosion, manufacturability and Tritium Breeding Ratio in normal condition such as a coolant velocity ≤8m/s, pipe diameter ≥5mm and a total FW thickness ≤22mm.However it is important to note that the FW channels currently fulfill additional functions and are therefore not optimized “at all cost” regarding heat load capacity and the paper points out some recommendations against missing assumptions.

The design of the ITER first wall panels

The ITER blanket is in the final stage of design completion. The issues raised during the 2007 ITER design review about the first wall (FW) heat loads and remote handling strategy have been addressed, while integrating the recently confirmed in-vessel coils. This paper focuses on the FW design, which is nearing completion. Key design justifications are presented, followed by a summary of the current status of the manufacturing plan and R&D activities.

Thermal analysis of new ITER FW and divertor design during VDE energy deposition

Further developments and investigations in the area of fusion energy devices require extensive and comprehensive computer simulations with great precision to evaluate reactor components behavior during normal and transient events. In this work we performed detailed study of the first wall (FW) subjected to high heat flux during a vertical displacement event (VDE) with various initial steady-state conditions and heat flux histories for the transient plasma energy deposition. We calculated the spatial temperature profile through out the entire module and the maximum surface temperature, as well as melting and vaporization thickness of Be surface during VDE and just before thermal quench (TQ). We further studied possible potential damage to plasma facing components (PFC) and structural materials for VDEs with higher energy loads than currently estimated.

Russian development of enhanced heat flux technologies for ITER first wall

Recently the ITER first wall (FW) design has been significantly upgraded to improve resistance to electromagnetic loads, to facilitate FW panel replacement and to increase FW ability to withstand higher (up to 5MW/m2) surface heat loads. The latter has made it necessary to re-employ technologies previously developed for the now-abandoned port limiters. These solutions are related to the cooling channel with CuCrZr–SS bimetallic walls and hypervapotron type cooling regime, optimization of Be-tiles dimensions and Be to CuCrZr joining technique. A number of representative mockups were tested at high heat flux (HHF) at the Tsefey electron-beam facility to verify the thermo-hydraulic characteristics of the reference cooling channel design at moderate water flow velocities (V=1–3m/s, P=2–3MPa, T=110–170°C). The heat flux was gradually varied in the range of 1–10MW/m2 until the critical heat flux was registered. The mockups of hypervapotron structure demonstrated the required cooling efficiency and critical heat flux margin (1.4) at a water velocity of ≥2m/s. Dimensions of Be armor tiles strongly affect the thermo-mechanical stresses both in the CuCrZr cooling wall and at the Be–CuCrZr interface. Results of tile dimensions optimization (variable in the range 12mm×12mm×6 to 50mm×50mm×8mm) obtained by the HHF (variable in the range of 3–8MW/m2) experiments are presented and compared with analysis. It is shown that optimization of the tile geometry and joining technology provides the required cyclic fatigue lifetime of the reference FW design.

Conceptual design of the liquid metal laboratory of the TECHNOFUSION facility

The application of liquid metal technology in fusion devices requires R&D related to many phenomena: interaction between liquid metals and structural material as corrosion, erosion and passivation techniques; magneto-hydrodynamics; free surface fluid-dynamics and any other physical aspect that will be needed for their safe reliable operation. In particular, there is a significant shortage of experimental facilities dedicated to the development of the lithium technology. In the framework of the TECHNOFUSION project, an experimental laboratory devoted to the lithium technology development is proposed, in order to shed some light in the path to IFMIF and the design of chamber's first wall and divertors. The conceptual design foresee a development in two stages, the first one consisting on a material testing loop. The second stage proposes the construction of a mock-up of the IFMIF target that will allow to assess the behaviour of a free-surface lithium target under vacuum conditions. In this paper, such conceptual design is addressed.

Experiment and analysis of hypervapotron mock-ups for preparing the 2nd qualification of the ITER blanket first wall

According to the increased heat flux condition up to 5 MW/m 2 in the International Thermonuclear Experimental Reactor (ITER), new design of the blanket first wall (FW) has been considered and the analysis was performed with ANSYS-CFX for checking its temperature with the ITER operation conditions. And a semi-prototype of the FW was proposed to be tested with the similar heat flux conditions under the second qualification for the FW procurement. In order to investigate the fabrication procedure and analysis capability of the code, two types of mock-up were fabricated according to the current semi-prototype design except for bending shape; one with hypervapotron and another without it. They were tested with KoHLT-2 (Korea Heat Load Test) facility and the results were compared with the ones by CFX code. The mass flow rate of inlet coolant was the same as the ITER condition and heat flux was loaded up to 0.48 MW/m 2 heat flux. The results show that the temperature of the mock-up can be predicted using the CFX code even with the complex geometry and the hypervapotron shows its function to increase the cooling.

Heat loads and shape design of the ITER first wall

A concept for a shaped first wall for ITER is presented. While keeping most features of the 2004 FDR wall (modules segmentation, plasma facing components technologies, plasma facing material), this concept provides protection of the lateral faces of the first wall panels against the intense parallel heat flux coming from the plasma. Excessive beryllium temperatures at the panel edges are avoided during regular operation. The intense heat flux at the top of the vessel is accounted for and protection is provided against the shine thru heat flux. Start-up and ramp down using the wall as a limiter is possible for up to 7.5 MW, both inboard and outboard. This is rendered possible by the use of 5 MW/m 2 technology panels for 40% of the panels.

Effect of cooling rate on mechanical properties of aged ITER-grade CuCrZr

The effects of the cooling rate on the microstructure and the mechanical properties of aged CuCrZr alloys were examined. The CuCrZr alloys were solution annealed at 980 °C for 0.5 h, cooled down by four different cooling methods and then aged at 400–620 °C for 2 h. The strength showed the highest value at an aging temperature of 440 °C irrespective of the cooling rate and it was decreased gradually with an increase of the aging temperature up to 620 °C. The strength of the CuCrZr alloys was highly dependant on the distribution of the precipitates and the highest strength was obtained in the CuCrZr alloys with a number of Guinier–Preston zones. For the same aging temperature, the strength was increased with the cooling rate but the difference in the strength became smaller when the CuCrZr alloys were aged at above 580 °C. The yield strength of the CuCrZr alloys was higher than the minimum value (150 MPa at 250 °C) required for the design of the International Thermonuclear Experimental Reactor first wall up to an aging temperature of 600 °C when a gas cooling method after a solution annealing was performed with cooling rates of 89 and 207 °C/min.

Development of a computational tool for limiter edge plasma modeling with application to IGNITOR

IGNITOR is the only experiment today designed with the ambitious goal of achieving ignition. During the discharge, the First-Wall Limiter (FWL) will receive about 20MW [B. Coppi, A. Airoldi, F. Bombarda, et al., Nucl. Fusion 41 (2001) 1253–1257] power, split between conduction/convection and radiation channels. Previous estimates suggested a peak heat flux on the FWL close to 1MW/m2. This value should be re-evaluated, accounting for the latest developments in the design of the IGNITOR first wall and changes in the operational scenarios, in order to assess the risk of damages to the FWL deriving from the combination of thermal and mechanical stresses. For this purpose, the development of the ASPOEL code, implementing a 2D model of the edge plasma and including a detailed representation of the FWL, was started at Politecnico di Torino. Here we present the main features of the new code, and illustrate its potential via a preliminary application to IGNITOR.