Julio Guirao

The European Procurement of Cold Test and Case Insertion of the ITER Toroidal Field Coils

The International Thermonuclear Experimental Reactor is an international scientific project with the aim of building a tokamak fusion reactor capable of producing at least 10 times more energy than that spent to sustain the reaction. In a tokamak the fusion reaction is magnetically confined and the toroidal field coil system plays a primary role in this confinement. Fusion for Energy, the European Domestic Agency for ITER, is responsible for the supply of 10 out the 19 toroidal field coils. Their procurement has been subdivided in three main work packages: the production of 70 radial plates (the structural components which will house the conductors), the manufacture of 10 winding packs (the core of the magnet) and cold test and insertion into the coil cases of 10 winding packs. The cold test/insertion work package presents significant technological challenges. These include the cold test of the winding packs 14 m high, 9 m wide and weighing 110 t, the welding and inspection of the 316 LN stainless steel coil case, with welded thicknesses of up to 144 mm accessible only from one side combined with the need to minimize the deformation during the welding process (more than 70 m of weld per coil and up to 90 passes to fill the chamfer) and the resin filling of the coil case after insertion of the winding pack (the total volume to be filled up is about one cubic meter per coil). From 2009 up to mid 2011, F4E has carried out an R&D program in order to investigate the most challenging steps of the manufacturing processes associated to this work package, both to meet the demands of the ITER schedule and to minimize technological risks; in this paper an overview of the results obtained is presented.

Electromagnetic analysis of ITER diagnostic port plugs and diagnostic components during plasma events

Eddy current induced electromagnetic (EM) loads during plasma events are the design driver for ITER port plug (PP) structure, diagnostic first wall (DFW), shield module (DSM) and diagnostic system supported by the PP structure. Generic models using commercial software OPERA, MAXWELL and ANSYS are developed and benchmarks are performed for global EM analysis to obtain port-specific design driving EM loads. The 20 degree vessel sector models of the upper and equatorial PP structure take the same ITER TF, CS and PF coil and plasma current (15 MA baseline plasma scenario) as input, then solve for eddy current induced on all passive structural components for various DINA disruption cases. The worst load case can be exponential decay or linear decay of plasma current depending on the dimension and location of diagnostic components inside the PPs. Static and transient magnetic fields from generic models are mapped onto an excel datasheet to establish component design load specification. Three levels of modeling effort are suggested. The global model analysis is used to validate the impact of component design to the PP global system response as well as to study its component dynamic effect as a result of the disruption loads on the PP structure. The local sub-model analysis can be used to extract more accurate EM loads on diagnostics and the excel datasheet of static and transient field maps are used as the initial design load specification for in-port components.

Integration of diagnostics on ITER

Diagnostics play a very important role in the modern Tokamak where optimum performance is essential. To achieve this, the device must be equipped with reliable and robust sensors and instrumentation that allow the operation envelope to be fully explored. Development of these diagnostics to maintain this reliability is necessary. Further to the development, the systems must be integrated in a way that maintains their performance while simultaneously satisfying the key requirements needed for safety and tokamak operation. ITER will have 50 diagnostics; almost all of which are utilized primarily for the real-time operation of the tokamak. While there is still much work to do, to date, significant progress has been made in the development of these systems. The work load for the developments is shared across all the ITER partners. This paper focuses on the challenges for the integration of the systems.

Environmental conditions & loads of ITER diagnostic equipment in the port plug interspace & port cell

Development of the port-based diagnostic and service system integration in the Tokamak has been a challenging task for ITER (Nuclear Facility INB-174) engineering. Port integration is demanding more work on ex-vessel to identify electromagnetic thermal and nuclear environment in interspace and port cell regions. Protecting the diagnostic equipment in the interspace & port cell regions poses considerable engineering constraints. This paper will focus on the normal and accidental environmental conditions such as seismic, magnetic field, radiation, pressure, temperature etc., with particular interest on US Domestic Agencies (USDA) based diagnostic systems.

Validation of Residual Stresses of Finite Element Simulation of Multi Pass Butt-Welded Plates Using the Contour Method

In this paper a comparison between the results obtained using multi pass welding finite element (FE) simulation and the contour method was made to evaluate the accuracy in residual stresses simulated for plates with different thicknesses. The contour method has been used to measure the residual stresses in multi pass butt-welded plates. Two 316 austenitic stainless steel multi pass Metal Inert Gas (MIG) butt-welded plates of 10 mm thickness were cut using wire Electric Discharge Machining (EDM). The measurements of the cross-section were made with a coordinate measuring machine (CMM) and the points obtained were used to calculate the residual stresses by mean of static analysis of finite elements. A multi pass welding FE simulation of the two plates was made to obtain the residual stresses after time cooling. The simulated results are generally in good agreement with the experimental measurements. Other plates of 25 mm thickness and the same material were multi pass MIG butt-welded to evaluate the behavior with different thicknesses. In this case the number of passes was 11. The same method was applied to obtain the residual stresses. A comparison between different thicknesses was made. The residual stresses validation will allow the finite element simulation to be used for the later simulation of residual stresses relaxation.

Development of Codes and Standards for ITER In-Vessel Components

This paper describes the changes made to existing version of the Structural Design Criteria for In-vessel Components (SDC-IC) within the ITER project, as a result of the revision and update process carried out recently. Several ITER components, referred to as In-vessel Components, are located inside the ITER Vacuum Vessel: (a) Blanket System: shields the Vessel and Magnets from heat and neutron fluxes; (b) Divertor: extracts heat, helium ash and impurities from the plasma; (c) Fuelling: gas injection system to introduce fuel into the Vacuum Vessel; (d) Ion Cyclotron Heating & Current Drive System: transfers energy to the plasma by electromagnetic radiation; (e) Electron Cyclotron Heating & Current Drive System: uses radio waves to heat to the plasma; (f) Neutral Beam Heating & Current Drive System: accelerates Deuterium particles into the plasma; (g) Lower Hybrid Heating & Current Drive System: drives electric current into the plasma; (h) Diagnostics: measurement systems to control plasma performance, and further understand plasma physics; (i) Test Blankets: demonstrate techniques for ensuring tritium production within the tokamak. ITER In-vessel Components will be subjected to special operating and environmental conditions (neutron radiation, high heat fluxes, electromagnetic forces, etc.). The effects of irradiation on them, including embrittlement, swelling and creep, are not addressed in the existing commercial codes. These conditions are different from conditions in fission reactors and create challenging issues related to the design of these components. For this reason the Structural Design Criteria for ITER In-vessel Components (SDC-IC) [1] was developed for design purposes. SDC-IC was based mainly on the RCC-MR [2] code, and included rules for assessment of effect of neutron irradiation. In 2008 some issues were identified: (1) Some parts had not been fully prepared to cover all needed areas for design; (2) Some important topics needed to be improved; (3) New editions of codes on pressure equipment had been published; (4) No manufacturing rules were included, so consistency between manufacturing rules to be used and design rules in SDC-IC needed to be demonstrated; (5) Compliance with the ESP (French Decree concerning the Pressure Equipment Directive 97/23/EC for non-nuclear pressure vessels) [3] and ESPN (French Order applicable for pressure vessels intended for nuclear facilities) [4] needed to be addressed. The work carried out for Fusion For Energy (European Union’s Joint Undertaking for ITER) is: (a) Modification of design rules, incorporating rules from recently developed codes, and development of specific design rules to cover ITER specific issues and operational conditions; (b) Demonstration of consistency between design rules in SDC-IC and european standards used for manufacturing, in particular EN 13445 [5]; identifying areas where consistency is not provided; (c) Assessment of the compliance with the Essential Safety Requirements of the French Regulations (ESP and ESPN).Copyright