First Wall Panels Research Articles

Multi-machine scaling of the main SOL parallel heat flux width in tokamak limiter plasmas

As in many of today’s tokamaks, plasma start-up in ITER will be performed in limiter configuration on either the inner or outer midplane first wall (FW). The massive, beryllium armored ITER FW panels are toroidally shaped to protect panel-to-panel misalignments, increasing the deposited power flux density compared with a purely cylindrical surface. The chosen shaping should thus be optimized for a given radial profile of parallel heat flux, in the scrape-off layer (SOL) to ensure optimal power spreading. For plasmas limited on the outer wall in tokamaks, this profile is commonly observed to decay exponentially as , or, for inner wall limiter plasmas with the double exponential decay comprising a sharp near-SOL feature and a broader main SOL width, . The initial choice of , which is critical in ensuring that current ramp-up or down will be possible as planned in the ITER scenario design, was made on the basis of an extremely restricted L-mode divertor dataset, using infra-red thermography measurements on the outer divertor target to extrapolate to a heat flux width at the main plasma midplane. This unsatisfactory situation has now been significantly improved by a dedicated multi-machine ohmic and L-mode limiter plasma study, conducted under the auspices of the International Tokamak Physics Activity, involving 11 tokamaks covering a wide parameter range with Measurements of in the database are made exclusively on all devices using a variety of fast reciprocating Langmuir probes entering the plasma at a variety of poloidal locations, but with the majority being on the low field side. Statistical analysis of the database reveals nine reasonable engineering and dimensionless scalings. All yield, however, similar predicted values of mapped to the outside midplane. The engineering scaling with the highest statistical significance, , dependent on input power density, aspect ratio and elongation, yields = [7, 4, 5] cm for = [2.5, 5.0, 7.5] MA, the three reference limiter plasma currents specified in the ITER heat and nuclear load specifications. Mapped to the inboard midplane, the worst case (7.5 MA) corresponds to mm, thus consolidating the 50 mm width used to optimize the FW panel toroidal shape.

Overview of JSC “NIKIET” activity on ITER Procurement Arrangements

The two following ITER blanket-relevant Procurement Arrangements (PA) were signed by Russian Federation and ITER Organization in 2014:1)1.6.P1ARF.01 “Blanket First Wall” (signed on 14-th of February, 2014);2)1.6.P3.RF.01 “Blanket Module Connections” (signed on 19-th of December, 2014).The first PA is devoted to the development, manufacturing, testing and procuring to ITER site of 179 Enhanced Heat Flux (EHF) First Wall (FW) Panels. These FW panels are intended to withstand the heat flux from plasma up to 4.7 MW/m2, and there are two institutions in Russian Federation responsible for the manufacturing, testing and delivering of these panels on the ITER site: JSC “NIIEFA” (Efremov Institute) and JSC “NIKIET”. JSC “NIIEFA” (Efremov Institute) will manufacture the plasma-facing components (PFC) of EHF FW Panels and perform the final assembling of the panels while JSC “NIKIET” will manufacture the FW beam structures, load-bearing structures of PFC and the all the elements of panel attachment system.As for the second PA (“Blanket Module Connectors”) the JSC “NIKIET” is the alone official Supplier and will manufacture and procure blanket flexible supports, electrical insulating key pads and shield block/vacuum vessel electrical connectors.This article briefly describes the joint activity of JSC “NIKIET” and Efremov Institute in the framework of 1.6.P1ARF.01 “Blanket First Wall” Procurement Arrangement and the material on the activity on the second PA. The main achievements on both PAs (during the period of 2014–2015) are presented and also critical issues and plans are underlined.

EU contribution to the procurement of the ITER blanket first wall

Fusion for Energy (F4E), the European Union’s Domestic Agency for ITER, is responsible for the procurement of about 50% of the ITER blanket first wall (FW), called normal heat flux FW. A procurement strategy has been implemented by the In-Vessel Project Team at F4E aimed at mitigating technical and commercial risks for the procurement of ITER blanket FW panels, promoting as far as possible competition among industrial partners. This procurement strategy has been supported by an extensive Research and Development (R&D) programme, implemented over more than 15 years in Europe, to develop various fabrication technologies. It includes in particular the manufacture and testing of small-scale, medium-scale mock-ups and full-scale prototypes of blanket FW panels. In this R&D programme, significant efforts have been devoted to the development of a reliable materials joining technique. Hot Isostatic Pressing was selected for the manufacture of the FW panels made from beryllium, copper–chromium–zirconium alloy and 316L(N)-IG austenitic stainless steel.This paper presents the main outcome of the on-going R&D programme, the latest results of the FW qualification programme together with the procurement strategy implemented by F4E for the supply of the European contribution to the procurement of the ITER blanket FW.

Progress on performance assessment of ITER enhanced heat flux first wall technology after neutron irradiation

ITER first wall (FW) panels are irradiated by energetic neutrons during the nuclear phase. Thus, an irradiation and high heat flux testing programme is undertaken by the ITER organization in order to evaluate the effects of neutron irradiation on the performance of enhanced heat flux (EHF) FW components. The test campaign includes neutron irradiation (up to 0.6–0.8 dpa at 200 °C–250 °C) of mock-ups that are representative of the final EHF FW panel design, followed by thermal fatigue tests (up to 4.7 MW m−2). Mock-ups were manufactured by the same manufacturing process as proposed for the series production. After a pre-irradiation thermal screening, eight mock-ups will be selected for the irradiation campaigns. This paper reports the preparatory work of HHF tests and neutron irradiation, assessment results as well as a brief description of mock-up manufacturing and inspection routes.

Thermo-mechanical analysis of the DEMO FW module

Thermomechanical performance of the first wall (FW) W/EUROFER sandwich type module is analyzed under DEMO reactor conditions. Engineering heat loads to the FW panels are estimated for steady state operation with the edge localized modes (ELMs). Calculations carried out by MEMOS code show the inhomogeneity of the material temperature due to discrete location of the water cooling tubes embedded into EUROFER. The hot spots are formed in the W armor and EUROFER between the cooling sectors and depend on the distance of their mutual locations. The bending stress due to vertical temperature gradients in W and EUROFER layers is calculated and remains smaller than the ultimate tensile stress for expected temperatures. Calculations show that under the Type I ELMs expected in DEMO the W surface melts at the ELMs peak positions and solidifies between ELMs. There is no temperature difference found between hot and cool spots during ELMs.

Current status of final design and R&D for ITER blanket shield blocks in Korea

The main function of the ITER blanket shield block (SB) is to provide nuclear shielding and support the first wall (FW) panel. It needs to accommodate all the components located on the vacuum vessel (in particular the in-vessel coils, blanket manifolds and the diagnostics). The conceptual, preliminary and final design reviews have been completed in the framework of the Blanket Integrated Product Team. The Korean Domestic Agency has successfully completed not only the final design activities, including thermo-hydraulic and thermo-mechanical analyses for SBs #2, #6, #8 and #16, but also the SB full scale prototype (FSP) pre-qualification program prior to issuing of the procurement agreement. SBs #2 and #6 are located at the in-board region of the tokamak. The pressure drop was less than 0.3 MPa and fully satisfied the design criteria. The thermo-mechanical stresses were also allowable even though the peak stresses occurred at nearby radial slit end holes, and their fatigue lives were evaluated over many more than 30 000 cycles. SB #8 is one of the most difficult modules to design, since this module will endure severe thermal loading not only from nuclear heating but also from plasma heat flux at uncovered regions by the FW. In order to resolve this design issue, the neutral beam shine-through module concept was applied to the FW uncovered region and it has been successfully verified as a possible design solution. SB #16 is located at the out-board central region of the tokamak. This module is under much higher nuclear loading than other modules and is covered by an enhanced heat flux FW panel. In the early design stage, many cooling headers on the front region were inserted to mitigate peak stresses near the access hole and radial slit end hole. However, the cooling headers on the front region needed to be removed in order to reduce the risk from cover welding during manufacturing. A few cooling headers now remain after efforts through several iterations to remove them and to optimize the cooling channels. The SB #8 FSP was manufactured and tested in accordance with the pre-qualification program based on the preliminary design, and related R&D activities were implemented to resolve the fabrication issues. This paper provides the current status of the final design and relevant R&D activities of the blanket SB.

Impact of a narrow limiter SOL heat flux channel on the ITER first wall panel shaping

The inboard limiters for ITER were initially designed on the assumption that the parallel heat flux density in the scrape-off layer (SOL) could be approximated by a single exponential with decay length λq. This assumption was found not to be adequate in 2012, when infra-red (IR) thermography measurements on the inner column during JET limiter discharges clearly revealed the presence of a narrow heat flux channel adjacent to the last closed flux surface. This near-SOL decay occurs with λq ∼ few mm, much shorter than the main SOL λq, and can raise the heat flux at the limiter apex a factor up to ∼4 above the value expected from a single, broader exponential. The original logarithmically shaped ITER inner wall first wall panels (FWPs) would be unsuited to handling the power loads produced by such a narrow feature. A multi-machine study involving the C-Mod, COMPASS, DIII-D and TCV tokamaks, employing inner wall IR measurements and/or inner wall reciprocating probes, was initiated to investigate the narrow limiter SOL heat flux channel. This paper describes the new results which have provided an experimental database for the narrow feature and presents an ITER inner wall FWP toroidal shape optimized for a double-exponential profile with λq = 4 (narrow feature) and 50 mm (main-SOL), the latter also derived from a separate multi-machine database constituted recently within the International Tokamak Physics Activity. It is shown that the new shape allows the power handling capability of the original shape design to be completely recovered for a wide variety of limiter start-up equilibria in the presence of a narrow feature, even taking assembly tolerances into account. It is, moreover, further shown that the new shape has the interesting property of both mitigating the impact of the narrow feature and resulting in only a very modest increase in heat load, compared to the current design, if the narrow feature is not eventually found on ITER.

Material migration studies with an ITER first wall panel proxy on EAST

The ITER beryllium (Be) first wall (FW) panels are shaped to protect leading edges between neighbouring panels arising from assembly tolerances. This departure from a perfectly cylindrical surface automatically leads to magnetically shadowed regions where eroded Be can be re-deposited, together with co-deposition of tritium fuel. To provide a benchmark for a series of erosion/re-deposition simulation studies performed for the ITER FW panels, dedicated experiments have been performed on the EAST tokamak using a specially designed, instrumented test limiter acting as a proxy for the FW panel geometry. Carbon coated molybdenum plates forming the limiter front surface were exposed to the outer midplane boundary plasma of helium discharges using the new Material and Plasma Evaluation System (MAPES). Net erosion and deposition patterns are estimated using ion beam analysis to measure the carbon layer thickness variation across the surface after exposure. The highest erosion of about 0.8 µm is found near the midplane, where the surface is closest to the plasma separatrix. No net deposition above the measurement detection limit was found on the proxy wall element, even in shadowed regions. The measured 2D surface erosion distribution has been modelled with the 3D Monte Carlo code ERO, using the local plasma parameter measurements together with a diffusive transport assumption. Excellent agreement between the experimentally observed net erosion and the modelled erosion profile has been obtained.

Modelling ELM heat flux deposition on the ITER main chamber wall

The interaction of ELM filaments with the ITER beryllium first wall panels (FWPs) is studied using a simple ad-hoc fluid model of the filament parallel transport, taking into account the full, three-dimensional structure of the FWPs, including magnetic shadowing effects. The calculated ELM surface heat loads are used as input to the RACLETTE heat transfer code to estimate the FWP surface temperature rise. The results indicate that controlled ELMs in ITER during burning plasma operation (ΔWELM≈0.6MJ) will not lead to melting or significant evaporation of the beryllium surfaces, even in the case of high ELM broadening and the minimum allowable distance between the primary and secondary separatrices. The ELM-averaged steady-state heat load also stays below the maximum power handling capability of the FWPs.

Prototyping of the Blanket Shield Module for the ITER EC H&CD Upper launcher

The design of the ITER Electron Cyclotron Heating and Current Drive (ECH&CD) Upper launcher is recently in the first of two final design phases. The first phase deals with the finalization of all FCS (First Confinement System) components as well as with specific design progress for the remaining In-vessel components.The most outstanding structural In-vessel component of an ECH&CD Upper launcher is the Blanket Shield Module (BSM) with the First Wall Panel (FWP). Both of them form the plasma facing part of the launcher, which has to meet strong demands on dissipation of nuclear heat loads and mechanical rigidity. Nuclear heat loads from 3MW/m3 at the First Wall Panel’ surface, decaying down to a tenth in a distance of 0.5m behind of it will affect the BSM and the FWP. Additional heating of maximum 0.5MW/m2 due to plasma radiation must be dissipated from the FWP.To guarantee save and homogenous removal of such extensive heat loads, the BSM is designed as a welded steel-case with specific cooling channels inside its wall structure. Attached to its face side is the FWP with a high-power cooling structure.Based on computational analysis the optimum cooling channel geometry has been investigated. Specific pre-prototype tests have been made and associated assembly parameters have been determined in order to identify optimum manufacturing processes and joining techniques, which guarantee a robust design with maximum geometrical accuracy.This paper describes the design, manufacturing and testing of a full-size mock-up of the BSM. The study was carried out in an industrial cooperation with MAN Diesel and Turbo SE.

CuCrZr alloy microstructure and mechanical properties after hot isostatic pressing bonding cycles

ITER first wall (FW) panels are a layered structure made of the three following materials: 316L(N) austenitic stainless steel, CuCrZr alloy and beryllium. Two hot isostatic pressing (HIP) cycles are included in the reference fabrication route to bond these materials together for the normal heat flux design supplied by the European Union (EU). This reference fabrication route ensures sufficiently good mechanical properties for the materials and joints, which fulfil the ITER mechanical specifications, but often results in a coarse grain size for the CuCrZr alloy, which is not favourable, especially, for the thermal creep properties of the FW panels. To limit the abnormal grain growth of CuCrZr and make the ITER FW fabrication route more reliable, a study began in 2010 in the EU in the frame of an ITER task agreement. Two material fabrication approaches have been investigated. The first one was dedicated to the fabrication of solid CuCrZr alloy in close collaboration with an industrial copper alloys manufacturer. The second approach investigated was the manufacturing of CuCrZr alloy using the powder metallurgy (PM) route and HIP consolidation. This paper presents the main mechanical and microstructural results associated with the two CuCrZr approaches mentioned above. The mechanical properties of solid CuCrZr, PM CuCrZr and joints (solid CuCrZr/solid CuCrZr and solid CuCrZr/316L(N) and PM CuCrZr/316L(N)) are also presented.

Preliminary Test and Evaluation of Nondestructive Examination for ITER First Wall Development in Korea

ITER first wall (FW) includes a beryllium armour joined to a Cu heat sink with a stainless steel back plate. These FW panels are one of the critical components in the ITER tokamak with a maximum surface heat flux of 5 MW/m2. Therefore, a qualification test needs to be performed with the goal to qualify the joining technologies required for the ITER FW. Various mock-ups were fabricated to develop the manufacturing procedure of FW components. For the nondestructive examination of the fabricated mock-ups, an ultrasonic test (UT) was performed with optimized probes. The UT was performed by using a three-axis digital ultrasonic C-scan system and software. The system comprised an ultrasonic pulser and receiver, model Panametrics 5800PR; a personal computer having an internal analog/digital converter board and four-axis motion control board; and a three-axis scanning tank. Two types of transducers were used for this experiment. One was Panametrics V312-SU, having a center frequency of 10 MHz (nominal) and a piezoelectric element diameter of 0.25 in with a flat protective layer for the Be/Cu. The other was Panametrics V309-SU with a center frequency of 5 MHz and an element diameter of 0.5 in for the Cu/SS interface. Winspect software controlled all aspects of data acquisition, motion control, data archiving, and image display. Based on the acceptance criteria, the average amplitude of the interface signals, which have about 50% of the reference echo amplitude, was recorded and analyzed on each beryllium tile. An image-analysis software analyzed the statistics of amplitude distribution and calculated the unacceptable area. Each mock-up that passed these UTs was concluded to qualify the joining technologies required for an ITER FW by using a high-heat flux test facility. As a result of these qualification tests based on the acceptance criteria of an ITER FW, the fabrication technologies will be utilized to develop the FW of plasma-facing components.

Overview of the design and R&D of the ITER blanket system

Abstract The ITER blanket design has substantially evolved since the ITER design review of 2007. Two major incentives for the design changes have been the need to account for large plasma heat fluxes to the First Wall (FW) and the need for acceptable maintenance of FW panels. In parallel to the design effort, a focused R&D program is being carried out including manufacturing and testing of semi-prototypes for the FW panels, and of full-scale prototypes for the shield blocks. This paper summarizes the status of the ITER blanket system design including the accommodation of interfaces, and describes some of the key R&D activities in support of the design with the goal of starting procurement in the first half of 2013.

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.

ERO code benchmarking of ITER first wall beryllium erosion/re-deposition against LIM predictions

Previous studies (Carpentier et al 2011 J. Nucl. Mater. 415 S165–S169) carried out with the LIM code of the ITER first wall (FW) on beryllium (Be) erosion, re-deposition and tritium retention by co-deposition under steady-state burning plasma conditions have shown that, depending on input plasma parameter assumptions and sputtering yields, the erosion lifetime and fuel retention on some parts of the FW can be a serious concern. The importance of the issue is such that a benchmark of this previous work is sought and has been provided by the ERO code (Pitts et al 2011 J. Nucl. Mater.415 S957–S964) simulations described in this paper. Provided that inputs to the codes are carefully matched, excellent agreement is found between the erosion/deposition profiles from both codes for a given ITER-shaped FW panel. Issues regarding the difficult problem of the correct treatment of Be sputtering are discussed in relation to the simulations. The possible influence of intrinsic Be impurity is investigated.

The upgraded laser in vessel viewing system (IVVS) for ITER

A prototype of a laser in vessel viewing and ranging system was developed at ENEA laboratories in Frascati, it uses the amplitude modulated laser radar concept and it is conceived to withstand the severe ITER conditions. The in vessel viewing system (IVVS) probe has been designed and built to perform sub-millimetric three-dimensional images inside ITER; it is based on an intrinsically radiation resistant concept and architecture. A first characterization of the IVVS probe under room conditions was made at Frascati Labs on a full-scale mock-up of the ITER first wall panel (FWP) and of the divertor vertical target. The first characterization demonstrated that an upgrade of the IVVS performances was necessary especially to cope with divertor surfaces made of carbon material, which is highly absorbent from optical point of view. The paper describes the new developments of IVVS prototype to increase range measurement performances that was obtained increasing the modulation frequency, the optical efficiency of the probe optics and the laser power. A new test campaign has been carried out on the upgraded IVVS and a compact characteristic curve describing its performances has been found both in mathematical and graphical form. As far as viewing is concerned, the system has confirmed the sub-millimetric viewing resolution, reaching in the worst cases ∼1 mm of resolution. The image quality was excellent in almost all the cases. The range measurement performance of IVVS system has been strongly upgraded reducing the standard deviation of range measure of a factor varying from 6 to 12. The increased performance allows measuring surface shapes and erosion on first wall tiles and divertor also for inclination angles completely outside the previous IVVS characteristics.

Manufacturing issues for the production of beryllium clad prototype components for ITER

This paper reviews the manufacture of beryllium clad plasma facing components by AMEC under the auspices of EFDA to support the development of ITER. Over the last 2 years, three types of component have been produced; thermal fatigue mock-ups, Plasma Wall Interaction (PWI) samples and a full-scale ITER First Wall Panel (FWP) prototype. In this paper the manufacturing stages used to produce the components are summarised and the non-destructive examination and heat flux test results from the work are presented. Two issues are highlighted that are relevant for the series production of FWPs for ITER. Specifically these are that an integrated manufacturing process, covering all manufacturing stages should be used to minimise cost and material wastage. Also that small variations in the component geometry can significantly affect the quality of the ultrasonic NDE scans used to assess the beryllium tile bond integrity during manufacture.

A conceptual design proposal for a blanket module in the neutral beam injector region of ITER

The development work carried out in Europe on the design of standard blanket shield modules for ITER has been continued, with the objective of further decreasing the fabrication and maintenance costs and demonstrating viable solutions for the most complicated special shield module of the neutral beam injector (NBI) region. A modified attachment system of first wall panels (FWP) to the shield block, using poloidal pins, has been proposed in order to select the same attachment technology for plasma facing components (PFC) of both blanket and divertor systems, to reduce the costs and to simplify the refurbishment of the PFC in the hot cell. The mechanical components of the new attachment system increase the engineering margins against electromagnetic loads and the modified layout of cooling circuits with the proposed fabrication technology decreases the risk of water leakage inside the vacuum vessel during operation. The complicated shape of the special modules of the NBI region, the higher thermal loads in this region and the expected higher probability for FWP exchange requires significant design changes compared to the standard modules developed and reported earlier. This paper describes a design concept developed for the NBI module No. 1 of the present ITER Blanket segmentation, with a proposed fabrication route. It shows clearly the great advantage of the proposed fabrication technology with respect to design flexibility and reliability compared to standard design and fabrication routes.

Design for in-pile thermal fatigue of first wall mock-ups under ITER relevant conditions

This paper describes the general design and the thermal analysis of an irradiation rig devoted to in-pile thermal fatigue testing of three actively cooled first wall (FW) mock-ups. The temperatures, stress distributions and stress amplitudes at the Be/CuCrZr interface of the mock-ups will be as close as possible to the values calculated for the international thermonuclear experimental reactor (ITER) FW panels. The heat flux will be provided by nuclear heating of high density (W) plates integrated on the Be-side. The assembly will be placed in the pool side facility of the high flux reactor in Petten. Thermal cycling is then arranged by mechanical movement towards and from the core box. As the thermal design of the irradiation rig is very critical, a pilot-irradiation will be performed to crosscheck the models used in the thermal and nuclear design of the rig.

Design and analysis of the structural components in the ITER ECH upper port plug

The structural components of the ITER ECH upper port plug are presented which integrates the mm-wave beam lines of the front steering launcher for the control of plasma instabilities. The design of the main structure considers two alternatives. Firstly, a double wall configuration for the port plug frame responds to homogeneous baking requirements at 240 °C. Secondly, an alternative frame is presented with a single wall in sections that do not require active cooling. The blanket shield module (BSM) is designed as a double-wall housing with a first wall panel (FWP) and individual shield blocks respecting the open space for the mm-wave beams. The cut-out sections to the design of the FWP and the neighbouring lower regular blanket module are defined for the beams of the actual front steering model. The thermo-mechanical stresses of the FWP and the BSM housing caused by nuclear heating and plasma radiation were analysed for the more severe case of the earlier remote steering (“3/8 RS”) reference model. As a result of the calculations for stationary loads and for transient loads during a typical plasma burn of 400 s, no critical stresses have to be expected in the BSM housing.