Blanket Shield Research Articles

The ITER EC H&CD upper launcher: Analysis of remote handling compatibility

The present design of the ECH (Electron Cyclotron Heating) upper port launcher has been evaluated in light of the ITER remote handling (RH) requirements. Changes to the launcher design associated with the accessibility, maintainability and manageability of replaceable components are presented. Captive bolts were placed along the flange of the Blanket Shielding Module (BSM). A hinge mechanism was integrated to simplify the (dis-)mounting of the BSM and a frame with incorporated cooling and actuation lines was suggested for simplified mounting and replacement of the steerable mirrors. Rotating the upper port plug upside-down improves maintenance access and component handling. Tools are proposed for manipulation of the port plug and its sub-components. The RH compatibility analysis can improve a design. Early consideration of RH requirements and implementation of necessary features is therefore vital.

Status of the EU DA electromagnetic analysis contribution to ITER

Fusion for energy (F4E), the European Domestic Agency for ITER, is involved in a relevant number of activities in the area of electromagnetic analysis in support of ITER general design, and of specific requirement of the EU in-kind procurement. In this context, its main activity is linked with the electromagnetic analysis of several ITER components (blanket shield modules and first wall panels, blanket cooling manifolds, TBM port plug, etc.) subjected to electro-dynamical loads. Another important activity is related to the ITER superconducting magnets, namely the quench detection of the ITER TF coils and the Joule losses in the magnets cold structures. Last, but not least, a further activity is on going on the error field analysis due to tolerances in both construction and assembly of ITER magnets.

Development of design options for the port plug components of the ITER core CXRS diagnostic

The paper presents component concepts developed for the ITER core charge exchange recombination spectroscopy (cCXRS). They are based on the cCXRS layout of 2009, named as reference design option. It includes an outer shell, carrying a blanket shield module and a shielding cassette. The cassette, attached to the outer shell, carries secondary mirrors and a retractable tube. The tube holds a first mirror (M1) and a shutter. Development priority is given to the M1, M1 holder, shutter, calibration system and retractable tube. These units are the most critical from reliability point of view. The developed design approaches could be likely used as generic or prototype solutions for a cCXRS diagnostic port plug in future. Before prototyping critical components as the M1 holder and shutter, the final cCXRS conceptual design has to meet forthcoming changes in the ITER diagnostic upper port plug (UPP) and blanket system. The integrated first wall—diagnostic shield modules, first wall recession, blanket shaping and generic UPP layout have considerable impact on the layout of the cCXRS and its components. The paper presents a preliminary solution for integration of the customized reference cCXRS into expected ITER blanket and UPP layouts.

Neutronic analysis of the ITER poloidal polarimeter

The nuclear heating rates of the optical mirrors of the poloidal polarimeter installed in the equatorial port plug of ITER are calculated. Since the system cannot have a sufficiently labyrinthine structure and the second mirrors are located closer to the plasma than the first mirrors due to limited space, the nuclear heating rate of the second mirrors is as high as that of the first mirrors. However, it is possible to reduce the nuclear heating rates more than 1 order if the blanket shield module provides a sufficient degree of neutron shielding. The effect of the neutron shield on the peripheral region of the upper port plug, in which the optical system of the poloidal polarimeter is installed, is also evaluated. The gamma-ray dose rate 10 6 s after shutdown of operations in the back flange region has been estimated to be ∼60 μSv/h when the neutron shield is filled in the port plug, which is below the design threshold for the ITER maintenance phase. Further, the nuclear heating rate of the Poloidal field coil near the same upper port has been evaluated to be more than 2 orders lower than the allowable value of 1 mW/cc.

Properties of unirradiated and irradiated Ti–6Al–4V alloy for ITER flexible connectors

The high strength (α+β) Ti–6Al–4V alloy was selected as the material for flexible attachments of the shield blanket modules in the ITER reactor. The different technologies used for manufacturing this alloy are: forging, stamping or pressing. The microstructures resulting from these processing methods can vary significantly and as a consequence the properties, including irradiation behavior, also vary. There are limited data available on the irradiation behavior of these materials. Specimens cut in the longitudinal and transversal directions of forged and stamped material were studied, with some of the specimens hydrogen charged to ∼400ppm H2. In the unirradiated condition the forged alloy had slightly more ductility than the stamped alloy. The strength properties of both were practically the same. Neutron irradiation of these materials in the IVV-2M reactor reached doses of ∼0.2 and 0.3dpa at temperatures 240–260°C. Irradiation resulted in substantial hardening and significant decrease of the fracture toughness of specimens from both materials.

Overview of Fusion Nuclear Technology Activities at SWIP

The fusion nuclear technology (FNT) activities at SWIP are overviewed, which are including the design and R&D of the helium-cooled solid breeder test blanket module (TBM), development of Liquid metal blanket technology, structure and function materials research for ITER shielding and test blanket, ITER shielding blanket material development and the joining technologies and the R&D for ITER gravity support.

Eddy current induced electromagnetic loads on shield blankets during plasma disruptions in ITER: A benchmark exercise

According to recent updates of ITER shield blanket design, electromagnetic loads during the plasma disruption are being evaluated to verify the mechanical confidence and reliability. As a course of such evaluations, a benchmark activity for the electromagnetic analysis, coordinated by ITER Organization, is underway between ITER parties to compare the calculation results for disruption loads on the blankets. In this paper, we present calculation results for the electromagnetic loads on the simplified but practical model of ITER shield blankets with respect to six representative disruption scenarios of which ITER distributes simulation results based on the DINA code as a reference of the design and analysis. Commercial finite element method software, ANSYS/Emag™, was employed to evaluate the eddy current on the blanket modules with the 40° sector model for major conducting structure of the tokamak including double-walled vacuum vessel, triangular support, and vertical targets of divertors. An interface between ANSYS/Emag™ and plasma simulator was implemented with a conversion tool assigning the plasma current density on the ANSYS elements corresponding to the current filaments in DINA outputs. Discussions are made of the possible improvement of the blanket model taking more realistic blanket configuration into account at the cost of the moderate increase in computational time. A final remark is given of the possibility of incorporating halo currents into ANSYS disruption simulations, which are major sources of electromagnetic loads on in-vessel components including blankets.

Neutronics experiments on HCPB and HCLL TBM mock-ups in preparation of nuclear measurements in ITER

In support of the breeder blanket development program, the EU is conducting a dedicated neutronics R&D effort to provide the basis for the design of nuclear tests to be performed in ITER on the Test Blanket Modules (TBMs). It includes the development of computational tools comprising both Monte-Carlo and deterministic transport, sensitivity and uncertainty codes, the generation of high quality neutron cross-section and covariance data libraries. These are validated experimentally in view of their application in the ITER TBM and the DEMO design. To this purpose, two neutronics experiments have been carried out on mock-ups of both the Helium Cooled Pebble Bed (HCPB) and the Helium Cooled Lithium Lead (HCLL) variants of ITER TBMs, at 14-MeV neutron sources. Redundant experimental techniques have been used to measure the resulting tritium production rate and the neutron and gamma ray spectra which are needed to predict the blanket shielding performance, nuclear power production and all nuclear loads. The comparison of experiment and corresponding calculation is obtained with the associated uncertainty margin based on experimental as well as calculational uncertainties. At the same time, suitable nuclear measuring techniques for TBMs in ITER, in particular for the tritium production, are being developed, optimised, and tested in the mock-up experiments. In particular, the present paper summarizes the preliminary results of the latest of such experiments, i.e. the one conducted on a mock-up of the HCLL blanket. It describes also the dedicated neutronics R&D activities, including the efforts to specify the neutronics tests and objectives in TBMs in ITER, considering the various design concepts.

Manufacturing studies of structural components for the ITER EC upper launcher

To counteract plasma instabilities, electron cyclotron launchers will be installed in four of the ITER upper ports. An EC launcher consists of a structural system which accommodates the MM-wave-components and has to meet demands on precise alignment, sufficient removal of nuclear heat loads, mechanical strength and proper nuclear shielding. The structural system consists of the Blanket Shield Module (BSM) and the Mainframe. Depending on the expected heat loads, the launcher components are designed as double-wall or single-wall elements, connected by massive flanges. Double-wall segments feature narrow cooling gaps with stiffening ribs between stainless steel shells. Single-wall segments consist of welded stainless steel plates with substantial material thickness. To investigate industrial manufacturing routes, characteristic sections of the BSM and the Mainframe were addressed in detail. To identify the optimum manufacturing strategy for double-wall components, three different concepts, namely HIP (Hot Isostatic Pressing), brazing and machining were studied and a total of four mock-ups of double-wall components were produced. Also for single-wall components appropriate manufacturing routes were investigated, optimum production parameters were determined and typical segments were manufactured. These prototypes are under study at the FZK Launcher Handling Test facility (LHT) where various ITER operating conditions can be simulated. Analyses and test series related to feasibility, prevention of residual stresses, contour accuracy, thermo-hydraulic behavior and also economical aspects were performed.

3D assessment of nuclear heating, dose rate, and structural damage in the generic ITER diagnostics upper port plugs and adjacent magnet coils

The USITER Project is responsible for providing a generic design of 18 diagnostics upper port plugs (UPPs). These plugs are designed to provide effective nuclear shielding, adequate nuclear heat removal capability, and allowance for personnel accessibility shortly after shutdown. In addition, the design limits for other parameters should not be exceeded during ITER lifetime. This includes the accumulative DPA and helium production in the UPP structure and the vacuum vessel, fast neutron fluence reaching the neighboring TF and PF magnets, and the accumulative local heating in their coil casings. In the present work, we assess these performance characteristics in three generic upper port plug (GUPP) designs of the in-plug shield and labyrinths that span possible configurations ranging from a conservative to worst case. The cases considered are: (1) Case A: “Visible-IR Camera diagnostic” with an intact port-attached blanket shield module (BSM), and (2) Case B: vis/IR system where the port-based BSM is replaced by overlapping adjacent vessel-attached BSM and (3) Case C: a “Large Aperture diagnostic” that is loosely resembles ECH style shielding plugs and aperture. These cases were modeled in a 20-degree sector of ITER. The 3D FEM Discrete Ordinates Code, ATTILA, was used in the analysis along with FENDL2.1 data and the FORNAX activation file.

Preliminary Thermal and Hydraulic Analysis on the ITER Upper Diagnostic Port Plug During Normal Operation and Baking

A preliminary thermo-hydraulic analysis was performed on the ITER diagnostic upper port plug. Relevant thermal and hydraulic parameters, such as coolant pressure drop, maximum structure temperature and bake-out time, were calculated for normal operation and baking. The upper port plug considered is based on the preliminary generic structure design of Princeton Plasma Physics Laboratory and the Blanket Shield Module (BSM) developed in Europe. The diagnostic shield modules are modeled so that the Korean diagnostic procurement package, which includes Vacuum Ultra-Violet (VUV) spectrometer and neutron activation system, can be integrated. The analysis provides design inputs to optimize flow in the cooling channels of the plug. The conjugated heat transfer analysis for the port plug confirms that it is important to secure accurate nuclear heat and accurate electro-magnetic (EM) force for the design of the joining flange between the BSM and the main body. Thermal analysis shows that it will take ten hours for the port plug to reach the bake-out temperature (240°C), if the window plate is heated additionally from the rear side.

Welding Technology Development for the Fabrication of ITER Blanket Shield Block in Korea

The ITER blanket shield block will be fabricated by the conventional process based on drilling, milling and welding of forged stainless steel blocks. For the adaptation of conventional method, one of the most important tasks to be verified is to develop the electron beam welding (EBW)technologies for the block to block joint. In addition, after drilling and milling for cooling passages, plugging techniques satisfying the requirements for the welding section should be developed. Three joining technologies, the electron beam welding for thick blocks, joining of lids and plugs, and attachment of flow drivers, were investigated and the results are summarized in this paper.Electron beam welding parameters for 70 and/or 110 mm thickness blocks satisfying the requirements in accordance with ASME code were established. It was also found that slower welding speed could suppress the formation of voids and porosities in weld metal. From the welding test results of mock-ups, optimal concepts on the weld design for front header lids or plugs were suggested.

Design and technical status of the EU contribution to ITER

Europe is involved in the procurement of most of the high-technology items for the ITER device (e.g. parts of the superconducting Toroidal (TF) and Poloidal Field (PF) coils, the vacuum vessel (VV), the in-vessel components, the remote handling, the additional heating systems, the tritium plant and cryoplant and finally parts of the diagnostics). In many cases the technologies required to manufacture these components are well established, in others there is still ongoing design and R&D work to select and optimise the final design solutions and to consolidate the underlying technologies as, for example, in the areas of heating and current drive, plasma diagnostics, shield blanket and first wall, remote handling, etc. A design review has recently been conducted by the ITER Organisation, with the support of the Domestic Agencies (DAs) established by the countries participating to ITER, to address the remaining outstanding technical issues and understand the associated implications for design, machine performance, schedule and cost. This paper provides an update of the design and technical status of EU contributions to ITER.

The ITER radial neutron camera: An updated neutronic analysis

The radial neutron camera (RNC) will provide the spatial distribution and the total strength of the ITER neutron source (emissivity profile and fusion power) by means of collimated neutron measurements. Line-integrated neutron spectral measurements can also provide information on the ion temperature profile. The present design of the RNC consists of two collimating structures for a full coverage of the plasma: 36 collimated lines of sight (LOS) distributed in three different planes view the plasma core (ex-port system) and nine collimated LOS view the plasma edge (in-port system). The RNC design is based on the combined use of the MCNP Monte Carlo code and a software tool performing asymmetric Abel inversion of simulated measured neutron signals (MSST). Neutron and γ-ray transport calculations are performed with MCNP using a 3D RNC model to determine the signal/noise for each RNC channel and the spectra at the detectors. The MSST code is used to check the RNC compliance with the ITER measurement requirements for the neutron emissivity profile. In the present paper the improvement of the hard variance reduction technique applied to the MCNP neutron source (consisting in sampling neutrons only from plasma regions contributing to the detector signal) is presented and the following issues are analyzed: the possibility of reducing the length of the ex-port collimators (resulting in a significant reduction of the overall RNC dimension and weight); options for the reduction of the dose due to the neutron streaming through the RNC cut-outs in the blanket shielding module; the integration of a γ-ray detection system in the RNC by partially filling the collimators with a neutron absorbing material (LiH).

Structural Design and Force Analysis for Equatorial Diagnostic Port Plug of ITER

The structural design of equatorial diagnostic port plug for international thermonuclear experimental reactor (ITER) was discussed in this paper. The induced electromagnetic (EM) loads of port plug and blanket shielding module (BSM) were presented. The finite element analysis (FEA) and numerical calculation of the system were performed under the maximum loads. According to the American Society of Mechanical Engineers (ASME) criteria, the EM stress is far below the admissible failure boundary, which indicates that the system can resist the cooperative action of EM loads and device deadweight. The analysis could provide technical data and references for further analysis and optimized design of the system.

Seismic response analysis for equatorial diagnostic port plug of international thermonuclear experimental reactor

Modal analysis and seismic response analysis were carried out for the equatorial diagnostic port plug of international thermonuclear experimental reactor (ITER). The aim of the theoretical analysis is to verify structural strength and reliability of the device. The working condition includes one-dimensional seismic wave and two-dimensional seismic wave. Modal analysis of the device shows that primary vibration is inclined to occur in low-order modes. The horizontal (X-direction, Y-direction) maximum vibration appears at the first and the fourth eigen modes, with the natural frequency of 70.59 and 215.88 Hz respectively, and the vertical (Z-direction) primary vibration appears at the second eigen mode with the natural frequency of 82.85 Hz. According to the results of the finite element analysis (FEA) program, the weak portions of the device are distributed in the joint of port body with blanket shielding module (BSM) and inner side wall of ribbed plate for lifting flange, the maximum von Mises stress is 14.8 MPa with the Y-direction seismic wave. In accordance with the design criteria, the destructive effect is far below the failure boundary, and the structural reliability of the equatorial diagnostic port plug can meet the requirements of the design specifications.

Conceptual design of the ITER upper port plug for charge exchange diagnostic

A plug for the ITER core charge exchange recombination spectroscopy (core CXRS) is located in the upper port 3. It transfers the light emitted by interaction of plasma ions with the diagnostic neutral beam (DNB). The plug consists of a main shell, a shielding cassette and a retractable tube. The tube carries the first mirror (M1) with a shutter and allows multiple replacements of M1. A large uncertainty in the M1 lifetime has initiated a new design option with M1 located as far as possible from the first wall. A fast shutter has been proposed to enhance M1 protection. It operates consistently with a pulsed DNB and uses elastic bearings. Three-point attachment scheme of blanket shield modules (BSMs) has been suggested. It is capable to react electromagnetic loads caused by eddy- and halo-currents for any BSM option with a good design margin. The concern has been raised with regards to the forces caused by halo current and applied to the entire plug. Possible solutions are either to shunt and support the plug to the port in the frontal part, or re-attach all BSM in this belt from the plugs to the vacuum vessel.

Prototyping studies for the Blanket Shield Module of the ITER ECH Upper Port Plug

A team of European associations is planning to procure ECH launcher turnkey systems for MHD control in the ITER plasma. ECH launchers will be installed to four ports on the upper level of the ITER vacuum vessel (VV). The structural system of the launchers accommodates the mm-wave components, cooling devices and elements for nuclear shielding. Its main components are the Blanket Shield Module (BSM), including the plasma facing First Wall Panel (FWP) and the port plug mainframe. A removable flange connection between the BSM and the main frame provides access to the internals. Appropriate remote handling capability is also taken as a design requirement. The BSM with the flange connection will be exposed to substantial nuclear heat loads. The manufacturing of machined components requires complex shaping with small tolerances and good quality of the surfaces due to operation under vacuum conditions. For the BSM and the front segment of the main frame a rigid double wall structure with meandering rectangular cooling channels was designed and analysed to meet these requirements. To investigate industrial manufacturing routes, a typical single-piece sample was machined and the manufacturing process was evaluated. Further two prototypes of a characteristic section of the BSM were manufactured, using two different fabrication techniques. These are (a) Hot Isostatic Pressing (HIP), which combines the sintering of metal powder inside of welded capsules and diffusion welding of solid parts and (b) brazing of bent and machined individual parts. The prototypes are under study at the Launcher Handling Test facility (LHT) at FZK, which offers a water circuit to provide coolant with adjustable parameters, simulating different ITER operating conditions. Extensive test series were performed to validate underlying analysis related to homogenous temperature distribution, tolerable pressure drop within the cooling paths and removal of applied heat loads.

European effort towards consolidating the design and manufacture of the ITER in-vessel components

The design of the in-vessel components represents one of the greatest technical challenges of the ITER machine. Europe has always played a very important role in supporting the design of the divertor and its high heat flux components, the first-wall the shield blanket and in developing and testing essential technologies and strengthening reliable manufacturing processes. This paper highlights the main research and development activities in progress in the European Fusion Programme in preparation of the start of ITER construction, to qualify all the technologies to be used, in particular, for the manufacture of the ITER plasma facing components and to demonstrate their compliance with the design requirements. It also describes some of the most outstanding problems that are still at issue in the design especially of the first-wall and that must be resolved to achieve the performance and machine availability goals and minimize the associated technical risks. The design and R&D priorities which are foreseen to achieve these aims are briefly mentioned.

Nitrogen segregation and blister formation of 316LN austenitic steel during electron beam welding tests for ITER gravity supports

316LN has been widely applied in the design of ITER components, such as shield blanket and gravity supports, due to its excellent corrosion resistance and high strength. The behavior of nitrogen in this steel during welding is important for the mechanical properties of the components. In this study, a focused 150 kv high voltage electron beam with 300 mA beam current has been used to weld 316LN steel under vacuum condition. The microstructure and composition of the welding area were observed and analyzed. The influence of welding on the shock resistance and tensile strength at both room temperature and low temperature were examined. It was found that the mechanical properties are strongly related to the defects formed in the welding process.