Blanket Shield Research Articles

The ITER EC H&CD Upper Launcher: Analysis of vertical Remote Handling applied to the BSM maintenance

This paper deals with Remote Handling activities foreseen on the Blanket Shield Module, the plasma facing component of the ITER Electron Cyclotron Heating and Current Drive Upper Launcher. The maintenance configuration considered here is the Vertical Remote Handling, meaning gravity acting along the launcher radial axis. The plant, where the maintenance under consideration is occurring, is the Hot Cell Facility Work Cell.The study here reported has been carried out within the presently ongoing EFDA Goal Oriented Training program on Remote Handling (GOT-RH), which aims to support ITER activities. This document and its contents have to be considered as part of a more vast RAMI analysis to be developed within the GOT-RH, which aims to maximize the Electron Cyclotron Heating and Current Drive system availability.The Baseline CAD model of the Electron Cyclotron Heating and Current Drive Upper Launcher is currently in its preliminary design phase and does not provide enough details for developing a fully detailed maintenance strategy. Therefore, through a System Engineering approach, a set of assumptions was conceived on the launcher structure, as a basis for development of a Remote Handling strategy. Moreover, to compare different design solutions related to the possibility of integrating a quasi-optical component into the Blanket Shield Module, a Trade-Off was made, and its contents are shown here.The outcome of this System Engineering approach has been formalized into Task Definition Forms whose contents are reported here. The Remote Handling strategy presented in this work will be tested in the near future both through Virtual Reality simulations and through prototype experiments.

Preliminary design of the ITER ECH Upper Launcher

The design of the ITER electron cyclotron launchers recently reached the preliminary design level - the last major milestone before design finalization. The ITER ECH system contains 24 installed gyrotrons providing a maximum ECH injected power of 20MW through transmission lines towards the tokamak. There are two EC launcher types both using a front steering mirror; one equatorial launcher (EL) for plasma heating and four upper launchers (UL) for plasma mode stabilization (neoclassical tearing modes and the sawtooth instability). A wide steering angle range of the ULs allows focusing of the beam on magnetic islands which are expected on the rational magnetic flux surfaces q=1 (sawtooth instability), q=3/2 and q=2 (NTMs).In this paper the preliminary design of the ITER ECH UL is presented, including the optical system and the structural components. Highlights of the design include the torus CVD-diamond windows, the frictionless, front steering mechanism and the plasma facing blanket shield module (BSM). Numerical simulations as well as prototype tests are used to verify the design

Nuclear Analysis of ITER Equatorial EC Launcher

Nuclear analysis of the ITER equatorial EC launcher consisting of an unique blanket shield structure and a port plug installing millimeter (mm) wave components, neutron shields, cooling water lines, etc. has been carried out. The analysis results are used to determine heat and/or particle loads on its components and to evaluate the possibility of “hands-on maintainability” (personnel accessibility) to the launcher back-end. A significant radiation leak at the gaps between the port walls and port plug frame of the launcher was revealed. Another significant neutron leakage is through the port wall consisting of only stainless steel but without light isotopes such as water. The shut down dose rates was estimated at the port interspace behind the launcher at the same level of the required value of 100 μSv/h. This analysis offers the potential to modify the launchers shielding layout to minimize the above leakage and further reduce the shut down dose rates in the regions of personnel access.

Challenges of ITER diagnostic electrical services

Diagnostic electrical services provide the electrical infrastructure to serve diagnostic components installed on the ITER tokamak. This infrastructure is composed of cables, connectors, cable tails, looms, conduits and feedthroughs. The diagnostic services offer as well a shelter for various instrumentations – vacuum vessel (VV), blanket and divertor. The diagnostic sensors are located on the inner and outer VV wall, on blanket shield modules, divertor cassettes and in port plugs. They require electrical cabling to extract the measurement and, in some cases, to supply electrical power to the sensors. These cables run from the sensors to feedthroughs on the VV and the port interspace or cryostat.The design and integration of all components that are part of diagnostic electrical services is an important engineering activity that is being challenged by the multiple requirements and constraints which have to be satisfied while at the same time delivering the required diagnostic performance. The positioning of the components must correlate not only with their functional specifications but also with the design of the major ITER components. This is a particular challenge because not all systems have reached the same level of design maturity.This paper outlines the engineering challenges of ITER diagnostics electrical services. The environmental conditions inside the VV will have an important impact. Leading risks to these components include poor electrical contact at connectors, the effects of exposure to nuclear irradiation, such as material transmutation, heating, and generation of spurious electrical signals etc., failure due to electromagnetic forces and electrical interference due to the noisy environment. Last but not least are the challenges for confinement and vacuum requirements set up on electrical feedthroughs. It will focus as well on the design and structural assessment of all components, and their requirements. Besides the integration limitations, the loads are the main design driver.

Influence of the blanket shield modules geometry on the operation of the ITER ICRF antenna

Three-dimensional electromagnetic simulations of the ITER ICRF antenna have been recently performed with the commercial code CST Microwave Studio® (MWS) [1]. A detailed model imported from the CATIA® file has been considered: it includes the 24 straps array (CY3.1 geometry [2]) and the surrounding blanket shield modules. The transient solver in MWS has detected the presence of a very localized peak in the input impedance matrix at a frequency of approximately 51MHz in vacuum conditions. The presence of such a resonance in the ITER operating range of frequency is of concern and should be understood as previous analysis reported in [3] concluded that TEM and non-TEM modes are not expected in this frequency band as long as the antenna is grounded to the port at 1m back from the antenna front face. By using a simplified model of the geometry we demonstrate that the resonance is a consequence of the considered geometry of the blanket shield modules and in particular of the cavity at the back of the modules made of the module attachment and the port plug outer face. We show that the presence of such a cavity locally increases the coaxial line impedance and allows for a TEM mode in the band. This physical analysis is supported by a transmission line model where the system made of the antenna and its surrounding port is approximated by a rectangular coaxial line.

The ITER EC H&CD upper launcher: Design, analysis and testing of a bolted joint for the Blanket Shield Module

The final design of the structural system for the ITER EC H&CD upper launcher is in progress. Many design features of the preliminary design are under revision with the aim to achieve the built-to-print-status. This paper deals with design and analysis of a bolted joint for the Blanket Shield Module with special perspective on Remote Handling capability. The BSM of the ECH Launcher is attached to the Launcher Main Frame by a bolted joint conceived so that in the Hot Cell Facility, RH maintenance can be performed on internal components. The joint must be capable to resist very high Electro-Magnetic loads from disruptions, while it has to sustain substantial thermal cycling during operation. Thus the need for a rigid and reliable design is essential. Beside the set of pre-stressed bolts the flanges were therefore equipped with additional shear keys to divert radial moments away from the bolts. Main focus of the work performed was the mechanical design of the joint and the assessment of the structural integrity with respect to the loads applied and its capability for maintenance by RH procedures. To fulfill a major aspect of the RH requirements, the bolts were designed as “captive” in order to avoid their accidental removal from the joint. The captive bolt design is based on a concept that uses a dedicated spring ring, a standard spiral spring and a tensioning screw with two threads to secure the bolts in a form-locking stop. The final approval phase of the RH compatibility foresees simulations in Virtual Reality and physical tests on prototypes.

The ITER EC H&CD Upper Launcher: Maintenance concepts

Maintenance of the ITER EC H&CD Upper Launcher (UL) shall be performed through the use of Remote Handling (RH) in the ITER Hot Cell Facility (HCF). The UL design will have to be fully compliant with ITER RH maintenance requirements and the set of RH tooling and services available in the HCF. This paper describes the development of an overall maintenance strategy for the UL, starting from a listing of all conceivable maintenance operations, including hands-on tasks. Components for which design concepts are discussed in this paper are the Blanket Shield Module (BSM), the steering mirror (M4), the mid optics (M1, M2) and the waveguide (WG) feed-through plate. Aspects related to RH documentation, overall maintenance strategy and design concepts for optimizing the maintainability of the UL are presented.

Optimisation of hot isostatic pressing bonded SS/SS joints conditions for ITER blanket shield

In the engineering design activity of international thermonuclear experimental reactor (ITER), stainless steels are being considered as candidates materials for several module type structures. Hot isostatic pressing (HIP) technique is expected for the fabrication of these modules. Stainless steel powders are simultaneously consolidated as mono-material block or/and joined in bi-material module.This paper reviews the manufacturing stages, non-destructive examination and the developments of the HIP bonded joints of 316L SS (powder and solid) for application to the ITER shield blanket.It is well known that the powder surface oxidation negatively influences the impact toughness of raw material and joints consolidated by this way. In order to get acceptable mechanical properties of materials, a study on the effect of reducing the powder oxygen content has been launched. To evaluate susceptibility to the oxygen content of HIPed joint specimens, tensile and toughness tests have been performed.From this study, optimal conditions of HIP were fitted and the influence of oxygen was mastered to obtain good mechanical properties of the consolidated powder material as well as for HIPed junction.

R&D Activities Regarding ITER Blanket First Wall in Korea

Korea Domestic Agency (KO-DA) was responsible for the procurement of the ITER blanket modules 1, 2, and 6 in the original procurement allocation. According to the procurement reallocation of the blanket system, Korea will procure the blanket shield block in place of the blanket first wall. Nevertheless, several R&D activities in Korea have been performed including optimization of the hot isostatic pressing (HIP) bonding process between Be/CuCrZr and CuCrZr/SS, the nondestructive test method, fabrication feasibility study, high heat flux tests, and design analysis. Especially, KO-DA participated in the qualification program for the mock-up manufacture and high heat flux tests. Several mock-ups were fabricated and tested during the qualification program. The details of the mock-up manufacture and test results are described in this paper. Also, two heat flux facilities were installed based on the graphite heating, and a new electron beam heat flux facility will be built in the near future for the enhanced heat flux mock-up test. As well, some design analysis was performed to investigate the performance of the blanket first wall against thermo-mechanical loading. In this paper, the status of the R&D activities and the results of the qualification tests for KO mock-ups are reviewed.

Comparative study of laser and TIG welding for application to ITER blanket hydraulic connection

The shield blanket in ITER has an active cooling structure necessitating hydraulic connections to the cooling water manifold. To maintain or replace the blanket, welding the hydraulic connection by remote handling is necessary. Access for the welding is limited in a small hole in the first wall because of spatial constraints related to neutron and heat fluxes. A bore welding tool is required. Laser and TIG welding tools have been developed, and the welding conditions have been optimized for all position welding to horizontally located pipes. Obtained parameters such as the weld heat input, allowable misalignment, the lifetime of the tools, and the amounts of spatter and fumes, are assessed comparatively. Considering additional aspects related to the remote handling compatibility, for example, the optical fiber or power cable handling, the preferable welding technique for the blanket remote maintenance is discussed.

Cooling optimization for preliminary design of ITER blanket shield block

Since the recommendation of blanket redesign by 2007 ITER design review, the blanket system has been developed in the framework of Blanket Integrated Product Team (BIPT) composed mainly of ITER organization (IO) and procuring parties. After the completion of blanket Conceptual Design Review (CDR), the design teams were organized to efficiently implement the preliminary and detailed design in a timely manner. The blanket shield block is a bulk structure to provide the nuclear shielding; therefore it should be designed to accommodate the nuclear heating in parallel with the mitigation of electromagnetic loading by slit optimization. This paper briefly describes the cooling optimization for the preliminary design of ITER blanket shield block 01, 02, 08 and 16. Three-dimensional hydraulic and thermo-mechanical analyses for the preliminary design modules of hybrid structure with poloidal and radial cooling concepts are performed under the inductive operation as a representative loading condition. The pressure drop, heat transfer and coolant uniformity in cooling passages are investigated in detail. In addition, the manufacturability of a blanket shield block is also considered as the important design constraint in the cooling optimization. Our stress evaluation follows the relevant codes and standards outlined in the design protocol provided by the IO. This paper presents the analysis results, identifies issues on the preliminary configuration and makes suggestions on the design improvements.

Neutronics Investigations for Helical DEMO Reactor FFHR-d1

Radiation shielding and tritium breeding performances of the helical DEMO reactor FFHR-d1 have been investigated for the present proposed reactor component configuration. Since the core plasma position shifts to inboard side of the torus, the total thickness of the inboard breeding blanket and radiation shield of FFHR-d1 is limited to ∼70 cm. To simulate the geometric features of the helical reactor, three-dimensional neutronics calculation model consisted of ∼4,000 cells have been prepared for the neutron and gamma-ray transport calculations using the MCNP code. It has been confirmed that the present radiation shield configuration with WC (tungsten carbide) and FS (ferritic steel) + B4C layers would provide sufficient shielding performance for the helical coils. The tritium breeding ratio (TBR) of 1.08 has been obtained with a Flibe+Be/FS breeding blanket for the present component configuration of FFHR-d1. c

High Performance Parallel Monte Carlo Transport Computations for ITER Fusion Neutronics Applications

Large scale neutronics calculations for radiation safety and machine reliability are required to support design activities for the ITER fusion reactor which is currently in phase of construction. Its large size and complexity of diagnostics, control and heating systems and ports, and also channel penetrations inside the thick blanket shielding surrounding the 14 MeV D-T neutron source are essential challenges for neutronics calculations. In the ITER tokamak geometry, the Monte Carlo (MC) method is the preferred one for radiation transport calculations. Due to the independence of particle histories, their tracks can be processed in parallel. Parallel computations on high performance cluster computers substantially increase number of sampled particles and therefore allow reaching the desired statistical precision of the MC results using the MCNP5 code. The MCNP5 parallel performance was assessed on the HPC-FF supercomputer. Use of CAD-based approach with high spatial resolution improves systematic adequacy of the MC geometry modeling. These achievements are demonstrated on radiation transport calculations for designing the Blanket Shield Module and Auxiliary Shield of the ITER Electron Cyclotron Heating (ECH) upper launcher. The spatial distributions of nuclear heating were analysed by using the graphical representation of the MCNP5 mesh-tally results in 2D and 3D plots.

Integration of the ITER in Vessel Coil System

The operation of the ITER machine requires the implementation of two sets of coil systems installed inside the vessel – the edge-localized mode (ELM) coil system and the vertical stabilization (VS) coil system. The ELM coils generate resonant magnetic perturbations in order to reduce high power deposition in the divertor induced by ELM heating and can as an option be used to control moderately unstable resistive wall modes (RWM). The VS coils provide fast vertical stabilization of the plasma. There are three ELM coils in each 40 degrees vacuum vessel (VV) sector; one each in the lower, middle and upper segments for a total of twenty seven individually powered coils. ELM coils are 6-turn rectangular coils. There are two VS coils in the VV, in the lower and upper segments below and above the lower and upper ELM coils respectively. Each upper or lower VS coil is made with 4 turns independently fed for failure recovery in the event of a faulted turn. The In-Vessel Coils (IVCs) and feeders are placed under the blanket shield modules and manifolds and need to be compatible with them. An integrated design concept has been developed that provides for an integrated design of the IVCs and their feeders, the blanket manifolds and the blankets and their respective attachment features to the VV.

Electromagnetic Analysis of Forces and Torques on Selected Components of the ITER Blanket System due to Plasma Disruption

The ITER device is based on the tokamak concept of magnetic confinement in which the plasma is contained by the use of strong magnetic fields. The nearest structure to the plasma is the blanket system which provides shielding to the vacuum vessel and the superconducting magnets. There are potential abnormal operating environments where the plasma currents inside the tokamak are disrupted and induce eddy currents in the blanket (first wall and shield module). These currents interact with the large magnetic fields to produce forces in the blanket which could potentially cause mechanical failure in the first wall, shield module, or vacuum vessel. For this reason the design and qualification of the ITER blanket system requires appropriate high-fidelity electromagnetic simulations that capture the physics of these disruption scenarios.A number of different geometries will be discussed revealing the effect of different first wall designs and shield modules on the forces and torques experienced by these assemblies during plasma disruption.The key features of the modeling procedure will be presented including the plasma current modeling and geometric modeling of the first wall, shield modules, and vacuum vessel. The eddy current calculation is performed using the Opera-3d software.

Thermo-Hydraulic Performance Analysis for Conceptual Design of ITER Blanket Shield Block

Since the recommendation of blanket redesign by 2007 ITER design review, the blanket system has been developed in the framework of blanket integrated product team composed mainly of ITER organization and procuring parties. As a part of blanket conceptual design tasks, Korea domestic agency has supported the design analyses with respect to the hydraulic and thermal performance of the inboard blanket shield block.Three dimensional thermo-hydraulic and thermo-mechanical analyses of the inboard conceptual model with the poloidal cooling concept were performed. Two kinds of operation scenarios, inductive and non-inductive operations, were considered as representative loading conditions. The pressure drop, heat transfer and coolant uniformity in cooling passages were investigated in detail. The stress evaluation according to relevant code and standard was carried out and thermal bowing at flexible supports was also investigated. This paper presents the detailed analysis results, identifies issues on the conceptual configuration and makes suggestions on design improvements. In addition, this manuscript briefly describes about the complementary study such as the comparison of heat transfer coefficients calculated by empirical formula and CFD, and the effect of surface roughness inside the cooling channels.

Design of the ITER In-Vessel Coils

The ITER project is considering the inclusion of two sets of in-vessel coils, one to mitigate the effect of Edge Localized Modes (ELMs) and another to provide vertical stabilization (VS). The in-vessel location (behind the blanket shield modules, mounted to the vacuum vessel inner wall) presents special challenges in terms of nuclear radiation (˜3000 MGy) and temperature (100 °C vessel during operations, 200 °C during bakeout). Mineral insulated conductors are well suited to this environment but are not commercially available in the large cross section required. An R&D program is underway to demonstrate the production of mineral insulated (MgO or Spinel) hollow copper conductor with stainless steel jacketing needed for these coils. A preliminary design based on this conductor technology has been developed and is presented herein.

Re-Evaluation of the Neutronics Experiments for the ITER Shielding Blanket Using ENDF/B-VII, JEFF-3.1, and JENDL-3.3

MCNP calculations for a benchmark representative of the fusion blanket neutronics shielding in SINBAD (Shielding Integral Benchmark Archive Database) were performed by using the four nuclear data libraries, ENDF/B-VII, JEFF-3.1, JENDL-3.3, and FENDL-2.1. Neutron and gamma flux spectra were calculated at two positions in a mock-up of the ITER inboard shield system. The results were compared with each other and also compared with measured data from the neutronics experiments for the ITER shielding blanket. For neutron spectra calculations, it is noted that the MCNPX calculations using all libraries agree well with experiments at positions A and B. For gamma spectra calculations, an overall good agreement can be stated and the tendency of a slight underestimation with penetration depth is observed.

Electromagnetic transients simulation using a shell approach for ITER cCXRS upper port plug due to plasma vertical displacement events

Transient electromagnetic (EM) analysis is presented focusing on main components of the ITER core charge exchange recombination spectroscopy (cCXRS) port plug developed by Forschungszentrum Jülich (FZJ), ITER-NL and UKAEA(CCFE) in 2009. The cCXRS primary function is to transfer the light in the visible part of spectra emitted by interaction of the plasma ions with a diagnostic neutral beam. The TYPHOON software package has been used for the EM analysis. The code is dedicated for simulation of transient electromagnetic processes using a shell approach in the integral–differential formulation to represent conducting structures with a set of multi-connected shells arbitrary located in a space. The advantage of the shell approach is a higher flexibility in modelling detailed structures as compared with widely used 3D models. On the other hand, the shell approach requires ultimate care in modelling relatively thick structures. These issues are discussed in the paper. Two vertical displacement events (VDE) which seem to result in the largest EM loads on the main cCXRS components have been agreed with FZJ and simulated. Transient electromagnetic processes caused by different sources have been considered separately, and then superimposed to obtain the total solution. Three types of transient processes for each type of VDE have been analyzed: (1) due to variations of a toroidal plasma current, shape and position and due to variations of poloidal field coils (PFC) and central solenoid (CS) currents, (2) due to variations of the Halo current and (3) due to variations of a toroidal magnetic flux of plasma. The analysis covers two options for electrical contact between the main shell (MS) of the port plug and the blanket shield module (BSM). The results are supposed to be used for benchmarking with independent 3D EM models developed for the upper port plug.

The ITER EC H&CD upper launcher: Structural design

Four ITER EC H&CD (Electron Cyclotron Heating and Current Drive) Upper Launchers will be installed in the ITER Tokamak to counteract plasma instabilities by injection of up to 20 MW of millimeter-wave power at 170 GHz. Each Launcher features a structural system which is equipped with eight beam lines in a Front-Steering arrangement. The Launcher development has reached the status of a preliminary design, since the corresponding review meeting was held in November 2009 at the ITER site in Cadarache. All design work is performed by several EU associations being contracted by Fusion for Energy (F4E). The structural design of the Upper Launcher consists of three sub-components: First of all the Blanket Shield Module (BSM), which fills the gap between the regular blankets. The BSM dissipates about 80% of the nuclear heating and envelopes the front mirrors of the mm-wave system. Further the Launcher Mainframe, which provides a rigid structure for precise and secure integration of the mm-wave system to guarantee reliable operation under all potential scenarios. Finally the internals, such as dedicated support structures for the mm-wave system, shielding elements and components for gas and coolant supply. The most challenging design aspects are proper dissipation of nuclear heating in zones of high heat flux, the mechanical integrity during plasma disruptions, the integration of sufficient shielding material and the precise alignment of the mm-wave system under tight space conditions. Furthermore the definition of efficient manufacturing routes with respect to tolerance compliance requires substantial investigation and, though the Launcher is designed for ITER lifetime, potential repair by adequate remote handling procedures must be considered. This paper presents the recent status of the preliminary structural design and outlines future design approaches with the main focus on manufacturing methods, remote handling capability of the sub-components and optimum integration of the internals to bring the EC Launcher towards the final design.