Blanket Shield Research Articles

Design modification of ITER equatorial EC launcher for electron cyclotron heating and current drive optimization

The design modification of the ITER equatorial electron cyclotron (EC) launcher to enhance off-axis current drive performance, changing the steering capability of injection beams from toroidal to poloidal, is described. The modification requirement impacts the design of all 3millimeter wave beam paths transmitted through the port plug and blanket shield modules (BSMs) installed in front. The optimized configuration of the mm wave design consisting of both steering and fixed mirrors, beam ducts between the waveguide outlet and BSM openings are successfully developed. Assuming a 100% pure HE11 mode propagated at the waveguide output, a 99% transmission efficiency from the waveguide outlet to the BSM output is obtained. This modification ensures that mm wave beams from both middle and bottom beam row pass through the same BSM opening and will lead to the increase of neutron shielding potential. The configurations of BSMs are also modified maintaining the structural design concept of each module. The preliminary result of electromagnetic analysis shows that induced force and momentum torque on the modified BSMs are below the mechanical criteria of the BSMs’ support structure.

Development of ITER Equatorial EC Launcher Components Toward the Final Design

Development of an electron cyclotron equatorial launcher has proceeded with a series of prototype tests and design enhancements intended to improve reliability and functionality of the launcher. The design enhancements include adaptation of the launcher steering angles such that one of three beam rows of the launcher is necessarily flipped to perform counter current drive to conform to a new ITER physics requirement. Also, the top and bottom steering rows have been tilted at an angle of 5 deg so that the top and bottom beam rows can be accessed from on-axis to near midradius. Furthermore, the position of the fixed focusing mirror that forms a quasi-optical in-vessel millimeter-wave (mm-wave) transmission line is modified to increase the nuclear shielding capability. A high-power experiment of the mm-wave launching system mock-up fabricated based on the design confirmed a successful steering capability of 20 to 40 deg. It was measured that some stray radio-frequency power propagated in the beam duct and generated some heat on the duct. Prototype tests also include the fabrication of the blanket shield module and partial port plug mock-up and have shown no serious technological issue regarding the fabrication and cooling functionality.

Manufacturing and testing of full scale prototype for ITER blanket shield block

Based on the preliminary design of the ITER blanket shield block (SB) #8, the full scale prototype (FSP) has been manufactured and tested in accordance with pre-qualification program, and related R&D was performed to resolve the technical issues of fabrication. The objective of the SB pre-qualification program is to demonstrate the acceptable manufacturing quality by successfully passing the formal test program.316L(N)-IG stainless steel forging blocks with 1.80L×1.12W×0.43t (m) were developed by using an electric arc furnace, and as a result, the material properties were satisfied with technical specification. In the course of applying conventional fabrication techniques such as cutting, milling, drilling and welding of the forged stainless steel block for the manufacturing of the SB #8 FSP, several technical problems have been addressed. And also, the hydraulic connector of cross-forged material re-melted by electro slag or vacuum arc requires the application of advanced joining techniques such as automatic bore TIG and friction welding. Many technical issues – drilling, welding, slitting, non-destructive test and so on – have been raised during manufacturing. Associated R&D including the computational simulation and coupon testing has been done in collaboration with relevant industries in order to resolve these engineering issues.This paper provides technical key issues and their possible resolutions addressed during the manufacture and formal test of the SB #8 FSP, and related R&D.

F4E studies for the electromagnetic analysis of ITER components

Fusion for Energy (F4E) is involved in a relevant number of activities in the area of electromagnetic analysis in support of ITER general design and EU in-kind procurement. In particular several ITER components (vacuum vessel, blanket shield modules and first wall panels, test blanket modules, ICRH antenna) are being analyzed from the electromagnetic point of view. In this paper we give an updated description of our main activities, highlighting the main assumptions, objectives, results and conclusions. The plasma instabilities we consider, typically disruptions and VDEs, can be both toroidally symmetric and asymmetric. This implies that, depending on the specific component and loading conditions, FE models we use span from a sector of 10 up to 360° of the ITER machine. The techniques for simulating the electromagnetic phenomena involved in a disruption and the postprocessing of the results to obtain the loads acting on the structures are described. Finally we summarize the typical loads applied to different components and give a critical view of the results.

Numerical Analyses of Electromagnetic Forces on the ITER Blanket Module Shield Block During Major Disruptions

Electromagnetic (EM) load is one of the key design drivers for the blanket shield block (SB) and other in-vessel components. In this article, an EM analysis method was developed to address the EM force on the SB. The plasma currents, which vary spatially and temporally, are loaded as a filament at each time point. The standard blanket module No.04 (BM04) under major disruption (MD) is selected to perform the analyses. The analyses results are validated by comparing currents on the passive structure. To better understand the effects of cooling channels and slits on the EM force, the case of SB without cooling channel and the case without slits are calculated to make comparisons. The results show that the slits play an important role in controlling the EM load on SB.

Design, performance, and grounding aspects of the International Thermonuclear Experimental Reactor ion cyclotron range of frequencies antenna

ITER's Ion Cyclotron Range of Frequencies (ICRF) system [Lamalle et al., Fusion Eng. Des. 88, 517–520 (2013)] comprises two antenna launchers designed by CYCLE (a consortium of European associations listed in the author affiliations above) on behalf of ITER Organisation (IO), each inserted as a Port Plug (PP) into one of ITER's Vacuum Vessel (VV) ports. Each launcher is an array of 4 toroidal by 6 poloidal RF current straps specified to couple up to 20 MW in total to the plasma in the frequency range of 40 to 55 MHz but limited to a maximum system voltage of 45 kV and limits on RF electric fields depending on their location and direction with respect to, respectively, the torus vacuum and the toroidal magnetic field. A crucial aspect of coupling ICRF power to plasmas is the knowledge of the plasma density profiles in the Scrape-Off Layer (SOL) and the location of the RF current straps with respect to the SOL. The launcher layout and details were optimized and its performance estimated for a worst case SOL provided by the IO. The paper summarizes the estimated performance obtained within the operational parameter space specified by IO. Aspects of the RF grounding of the whole antenna PP to the VV port and the effect of the voids between the PP and the Blanket Shielding Modules (BSM) surrounding the antenna front are discussed. These blanket modules, whose dimensions are of the order of the ICRF wavelengths, together with the clearance gaps between them will constitute a corrugated structure which will interact with the electromagnetic waves launched by ICRF antennas. The conditions in which the grooves constituted by the clearance gaps between the blanket modules can become resonant are studied. Simple analytical models and numerical simulations show that mushroom type structures (with larger gaps at the back than at the front) can bring down the resonance frequencies, which could lead to large voltages in the gaps between the blanket modules and perturb the RF properties of the antenna if they are in the ICRF operating range. The effect on the wave propagation along the wall structure, which is acting as a spatially periodic (toroidally and poloidally) corrugated structure, and hence constitutes a slow wave structure modifying the wall boundary condition, is examined.

Mechanical Performance of the ITER ELM Coils' Bracket Material of Inconel 625

The effects of edge localized modes (ELMs) in ITER Tokamak device are being designed to be addressed by the ELM coils. The designed ELM coils include nine toroidal sectors of three (upper, midplane, and lower) six-turn stainless steel jacketed mineral insulated coils. They will be mounted to the vacuum vessel wall behind the blanket shield modules, and fixed by a series of brackets. Mechanical performance of the brackets should achieve the design requirements to resist with the special challenges, such as the nuclear radiation (~ 3000 MGy), high temperature (~ 200°C), as well as bearing the additional loads from ELM coils. Inconel 625 is designed to manufacture the ITER ELM coils' brackets. This paper focuses on the mechanical performance of Inconel 625 materials after forging and heat treatment, especially the fatigue performances. The tensile properties and the fatigue crack growth rate of Inconel 625 after forging and heat treatment have been tested at room temperature. The test results show that the mechanical properties of the Inconel 625 material fit the design requirement.

Joining of HHF materials applying electroplating technology

Tungsten will be used as armor material for blanket shielding and is designated as high heat flux material for divertors, beyond application of improved W composite alloys as structural material. Independent from design (water- or helium-cooled), a successful development is inherently correlated with joining of tungsten with functional components.Depending on the design variants, the fabricated joints have to guarantee specific functional or structural properties, e.g., good thermal conductivity or mechanical load transmission. Tungsten shows lacks in adapted joining due to its metallurgical behavior ranging from immiscibility over bad wetting to brittle intermetallic phase formation. Electroplating has shown to overcome such drawbacks and that homogeneous functional (e.g., Ni or Pd) and filler (e.g., Cu) layers can be deposited.In this paper the progress achieved in development of electroplating processes for joining W to W or steel to steel will be shown. The main focus will be the characterization of the processed joints applying metallurgical investigations including SEM/EDX analyses and non-destructive testing. The mechanical stability of the produced joints is demonstrated by presenting recent shear test data. The W–W joints failed due to cracking in W, whereas the steel–steel joints cracked in the brazing zone at about 200N/mm2 load.

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.

Multiaxial fatigue analysis for IMIC of ITER upper ELM coil

Inconel Jacketed Mineral Insulated Conductor (IMIC) is a very important component of International Thermonuclear Experimental Reactor (ITER) Edge Localized Modes (ELM) coils, which are located between the vacuum vessel (VV) and blanket shield modules and subject to high radiation levels, high temperature and high magnetic field. These coils will experience thermal pulsed, cyclic electromagnetic (EM) load during operation. They are designed to sustain at 1.5e8 total stress cycles and shall have sufficient strength and excellent fatigue to transport and bear the high cyclic load. For IMIC, multiaxial fatigue analysis is used to evaluate failure. Two methods based on the alternating stress and mean stress in American Society of Mechanical Engineers (ASME) code provide the design codes for multiaxial fatigue evaluation: constant principal stress direction and variation of principal stress direction. Results show that using the two methods obtains basically the same equivalent alternating stress. Both of them can be recommended for the ELM coils and IMIC can meet the fatigue criteria.

CFD analysis of the ITER first wall 06 panel. Part I: Model set-up and flow distribution

The first detailed Computational Fluid Dynamics (CFD) analysis of the FW06 panel of the ITER shielding blanket is presented in two companion papers. In this Part I we introduce the problem, define the model together with its input and discuss the results with particular reference to the hydraulics of the water coolant. The pressure drop across the panel is computed, together with the distribution of the flow among the different channels. Different design options are studied, with particular reference to the minimization of stagnation/recirculation regions.

CFD analysis of the ITER first wall 06 panel. Part II: Thermal-hydraulics

The computational fluid dynamics (CFD) analysis of the FW06 panel of the ITER shielding blanket is presented in two companion papers. In this Part II we concentrate on the thermal-hydraulics of the water coolant, driven by the nuclear volumetric and plasma surface heat loads discussed in Part I. Both the detailed steady state analysis of a single cooling channel and the coarse transient analysis of the whole panel are considered. The compatibility of the hot spots with the maximum recommended temperatures for the different materials is confirmed. The heat transfer coefficient between coolant and walls is obtained post-processing the results of the simulation and compared with the results of available correlations, which may be used for simpler analyses: in the fully developed flow regions of the cooling pipes, it turns out to be well approximated by the Sieder–Tate correlation. The operation margin with respect to the critical heat flux is also computed and turns out to be sufficiently large compared with the design limit.

Progress of the ECRH Upper Launcher design for ITER

The design of the ITER ECRH system provides 20MW millimeter wave power for central plasma heating and MHD stabilization. The system consists of an array of 24 gyrotrons with power supplies coupled to a set of transmission lines guiding the beams to the four upper and the equatorial launcher. The front steering upper launcher design described herein has passed successfully the preliminary design review, and it is presently in the final design stage. The launcher consists of a millimeter wave system and steering mechanism with neutron shielding integrated into an upper port plug with the plasma facing blanket shield module (in-vessel) and a set of ex-vessel waveguides connecting the launcher to the transmission lines.Part of the transmission lines are the ultra-low loss CVD torus diamond windows and a shutter valve, a miter bend section and the feedthroughs integrated in the plug closure plate. These components are connected by corrugated waveguides and form together the first confinement system (FCS). In-vessel, the millimeter-wave system includes a quasi-optical beam propagation system including four mirror sets and a front steering mirror. The millimeter wave system is integrated into a specifically optimized upper port plug providing structural stability to withstand plasma disruption forces and the high heat load from the plasma side with a dedicated blanket shield module. A recent update in the ITER interface definition has resulted in the recession of the upper port plug first wall panels, which is now integrated into the design. Apart from the millimeter wave system the upper port plug houses also a set of shield blocks which provide neutron shielding. An overview of the actual ITER ECRH Upper Launcher is given together with some highlights of the design.

Concept for a vertical maintenance remote handling system for multi module blanket segments in DEMO

The anticipated high neutron flux, and the consequent damage to plasma-facing components in DEMO, results in the need to regularly replace the tritium breeding and radiation shielding blanket. The current European multi module segment (MMS) blanket concept favours a less invasive small port entry maintenance system over large sector transport concepts, because of the reduced impact on other tokamak systems – particularly the magnetic coils. This paper presents a novel conceptual remote maintenance strategy for a Vertical Maintenance Scheme DEMO, incorporating substantiated designs for an in-vessel mover, to detach and attach the blanket segments, and cask-housed vertical maintenance devices to open and close access ports, cut and join service connections, and extract blanket segments from the vessel. In addition, a conceptual architectural model for DEMO was generated to capture functional and spatial interfaces between the remote maintenance equipment and other systems. Areas of further study are identified in order to comprehensively establish the feasibility of the proposed maintenance system.

Preliminary assessment for dust contamination of ITER In-Vessel Transporter

After plasma operations of ITER, radioactive dust will have accumulated in the vacuum vessel (VV). The In-Vessel Transporter (IVT) will be introduced into the VV to remove the shield blanket modules for maintenance or replacement and later reinstall them. The IVT itself also needs to undergo regular maintenance in the Hot Cell Facility (HCF). It is assumed that maintenance workers will be exposed to radioactive dust that has adhered to the surfaces of the IVT. In this study, the areas of the IVT that may be contaminated by dust are evaluated to assess the level of exposure to workers during maintenance work in the HCF. Decontamination processes for the IVT, such as a combination of vacuuming and brushing, were investigated and the dose rate after these processes was evaluated. Even though dust was removed from surfaces where decontamination was possible, the dose rate was very high at some assessment points. To decrease the dose rate in accordance with ALARA policy, a decontamination plan and a maintenance plan, which includes the removal of dust, a radiation shield system, and a reduction in working time are proposed.

Status of ITER TBM port plug conceptual design and analyses

The test blanket module port plug (TBM PP) consists of a TBM frame and two TBM-sets. However, at any time of the ITER operation, a TBM set can be replaced by a dummy TBM. The frame provides a standardized interface with the vacuum vessel (VV)/port structure and provides thermal isolation from the shield blanket. As one of the plasma-facing components, it shall withstand heat loads while at the same time provide adequate neutron shielding for the VV and magnet coils. The frame design shall provide a stable engineering solution to hold TBM-sets and also provide a mean for rapid remote handling replacement and refurbishment. This paper presents main design features of the conceptual design of TBM PP with two dummy TBMs. Also analysis results are summarized to evaluate shielding, hydraulic, and thermal and structural performances of the TBM PP design.

Heat transfer study for ITER blanket shield block cooling design

Two simplified models were developed for the cooling design of ITER shield block. Moreover, a new model, circular cylinder centered in a square solid, was also adopted to estimate the temperature, where the effects of heat transfer coefficient and volumetric heat rate were separated and studied individually. After that, the impact of dimension on the heat transfer in the new model was studied by a series of numerical analyses. At the last part, a numerical steady-state thermal analysis of a typical full shield block (SB) was performed to verify these models. Comparisons of the results from numerical analysis with these models show that the difference is acceptable in the practical application. The methods can be used not only for the cooling design, but also to know about the heat transfer in the SB.

Optimization of millimeter wave system in ITER Equatorial EC H&CD Launcher

A high power (20MW) and CW millimeter wave (mm-wave) injection is planned for Electron Cyclotron Heating and Current Drive (EC H&CD) in ITER. An optimization of the mm-wave system for the ITER EC H&CD Equatorial Launcher (EL) is performed. The optimization of the system is aimed to obtain the maximum transmission efficiency on the condition that 1.8MW injection per waveguide, ∼20cm in beam radius at the resonance layer and narrow opening of the Blanket Shielding Module (BSM). The transmission efficiency of 99.1% from the end of the waveguide inside the launcher to the output of the BSM is achieved.The mm-wave propagation with high order modes is also calculated by using an experimentally obtained high power mm-wave beam pattern that includes 95%HE11, 0.6%LP11, 0.2%LP02 and 4.2% other higher order modes. The analysis predicts the 1–2% additional loss will be induced by the high order modes.

Status of Design and R&D for ITER Blanket in Korea

Since the decision of blanket redesign by 2007 ITER design review, the blanket system is being developed in the framework of Blanket Integrated Product Team (BIPT) composed mainly of ITER Organization (IO) and procuring parties. Korean Domestic Agency (KODA) is mainly contributing to the design and development of blanket Shield Block (SB). In particular, KODA is supporting the design activities including electromagnetic, thermo-hydraulic and thermo-mechanical analyses to complete the final design of blanket shield block. For the manufacturing of a blanket shield block conventional fabrication techniques based on drilling, milling and welding of stainless steel forged blocks have been adopted. As a consequence of the manufacturing feasibility study, key fabrication techniques to be verified beforehand have been identified and successfully developed in collaboration with related industries. The pre-qualification program of the fabrication and testing of Full Scale Prototype (FSP) is in progress. Until now the material development of 316L(N)-IG stainless steel forging has been successfully completed, and the fabrication of FSP is on-going. Even though the procurement of blanket First Wall (FW) was withdrawn at the 9th meeting of the ITER Management Advisory Committee, the participation of the 2nd pre-qualification program of EHF (Enhanced Heat Flux) FW small scale mock is being valid for securing core engineering technologies. At present the fabricated mock-ups are waiting for high heat flux test with the Electron Beam (EB) gun test facility being newly built in Korea. This paper provides the current status of design and relevant R&D activities of the blanket system to secure key technologies and to fulfill our promise to ITER project.

Neutronics analysis of inboard shielding capability for a DEMO fusion reactor CFETR

The inboard shielding of a fusion reactor can be a crucial issue due to the limited space available in a tokamak configuration. It is necessary to assess the inboard shielding capability of DEMO for its initial design. In this paper, 1D and 3D neutronics models were developed based on a reference design of the Chinese Fusion Engineering Testing Reactor (CFETR). The neutron wall load (NWL) is in the range of 1.5–3MW/m2 and the inboard shielding thickness is constrained within 40–70cm in order to achieve the tritium self-sufficiency of the reactor. Referring to the detailed design of the ITER Toroidal Field Coils (TFCs) and using radiation hardening technology developed for ITER, the inboard blanket shielding capability and nuclear responses of the TFC are investigated for both FLiBe and Li4SiO4 breeding blanket concepts. The impact of the gaps on shielding performance is discussed. Some suggestions on improving the inboard shielding performance for DEMO are also proposed.