V. Barabash

Assessment of Tungsten for Use in the ITER Plasma Facing Components

Tungsten is one of the candidate armor materials for the plasma facing components of the International Thermonuclear Experimental Reactor (ITER). For the present reference design, tungsten has been selected as armor for the divertor upper vertical target, dome, cassette liner, and for lower baffle because of its unique resistance to ion and charge-exchange particle erosion in comparison with other materials. The issues related to the use of tungsten in ITER are described in this paper. The different tungsten grades (pure, dispersion strengthened and cast alloys) which are being considered as candidate materials are evaluated. A comparative analysis has been made of the mechanical properties of the various tungsten grades in different thermomechanical conditions, including the impact of irradiation effects. The different tungsten armor design solutions are also described.

Overview on fabrication and joining of plasma facing and high heat flux materials for ITER

This paper presents the results of the R&D program on the development of the joining technologies, including non-destructive inspection, for the high heat flux components carried out by the three ITER Participating Teams. Reliability requirements and design criteria are also discussed. The large amount of R&D performed within the ITER project has resulted in the development of suitable technologies, which meet or even exceed the design requirements and form a solid basis for the possible future construction of the ITER machine. However, further developments are still needed for the consolidation of the positive results with improving the reliability of the joints and reducing the manufacturing cost.

ITER full tungsten divertor qualification program and progress

The full tungsten divertor qualification program was defined for the R&D activity in domestic agencies. The qualification program consists of two steps: (i) technology development and validation and (ii) a full-scale demonstration. Small-scale mock-ups were manufactured in Japanese and European industries and delivered to the ITER divertor test facility in Russia for high heat flux testing. In parallel activity to the qualification program, both domestic agencies demonstrated that W monoblock technologies withstanding up to 20 MW m−2 were available.

Preliminary Safety Analysis of ITER

Abstract In order to support the licensing application for the ITER facility at Cadarache, a preliminary safety case has been prepared and submitted to the French nuclear safety authorities. This paper provides an overview of technical aspects of this case, which is based on an evolution of the safety approach developed and applied in earlier phases of the ITER project. The basis of the safety of ITER derives from the fundamental safety characteristics of fusion. The potential radiological hazards that arise are related to the tritium fuel and material activated by neutrons. The confinement of these materials is therefore the principal safety function, and it is reliably provided by robust barriers inherent in the design together with filtering and detritiation as a secondary level of confinement provision. A Defense in Depth approach is taken to ensure that off-normal events are minimized in their frequency, and that the consequences of accidents, even though extremely unlikely, are limited. A comprehensive set of analyses of postulated event sequences provides the demonstration that the consequences of enveloping scenarios are well within acceptable limits, and that even for hypothetical events involving two or more independent failures, the public and environmental impacts remain limited. An ALARA approach is taken to minimizing occupational radiation exposure, as well as other potential impacts of normal operation such as routine releases. Other hazards arising from internal and external risks are also considered, with design provisions, for example the Tokamak building is built on seismic isolation pads to minimise the effect of an earthquake.

Performance of the different tungsten grades under fusion relevant power loads

Abstract The test results of several W grades at conditions typical for plasma facing component operations are summarised. These include the effects of steady-state heat fluxes (up to 43 MW / m 2 ), disruption simulation (up to 30 MJ / m 2 during 0.05–0.36 ms) and heat flux tests of W after disruption simulation. Representatives of the main W grades have been investigated: pure sintered W, W–Re and W–Mo cast alloys, W –1% La 2 O 3 , W –2% CeO 2 , single crystal W, etc. The resistance to high heat fluxes strongly depends on the orientation of the W grains to incident heat flux and with proper orientation W can withstand heat fluxes up to 27 MW / m 2 . After disruption simulation, intensive surface crack formation has been observed for all studied W grades except single crystal W. Severe damage after disruption and thermal fatigue loading have been observed for almost all W grades except the W–5Re–0.1ZrC alloy and W–Re single crystal.

ITER vacuum vessel design and construction

Abstract According to recent design review results, the original reference vacuum vessel (VV) design was selected with a number of modifications including 3D shaping of the outboard inner shell. The VV load conditions were updated based on reviews of the plasma disruption and vertical displacement event (VDE) database. The lower port gussets have been reinforced based on structural analysis results, including non-linear buckling. Design of in-vessel coils for the mitigation of edge localized modes (ELM) and plasma vertical stabilization (VS) has progressed. Design of the in-wall-shielding (IWS) has progressed in details. The detailed layout of ferritic steel plates and borated steel plates is optimized based on the toroidal field ripple analysis. The procurement arrangements (PAs) for the VV including ports and IWS have been prepared or signed. Final design reviews were carried out to check readiness for the PA signature. The procedure for licensing the ITER VV according to the French Order on Nuclear Pressure Equipment (ESPN) has started and conformity assessment is being performed by an Agreed Notified Body (ANB). A VV design description document, VV load specification document, hazard and stress analysis reports and particular material appraisal were submitted according to the guideline and RCC-MR requirements.

Final design and start of manufacture of the ITER Vacuum Vessel ports

The ITER Vacuum Vessel (VV) features upper, equatorial and lower ports. Although the port design has been overall completed in the past, the design of some remaining interfaces was still in progress and has been finalized now. As the ITER construction phase has started, the procurement of the VV ports has been launched. The VV upper ports will be procured by the Russian Federation DA, while the equatorial and lower ports will be procured by the Korean DA. The main industrial suppliers were selected and development of the manufacturing design is in progress now. Since the VV is classified at nuclear level N2, design and manufacture of its components are to be compliant with the French code RCC-MR and regulations of nuclear pressure equipment in France. These regulations make a strong impact to the port design and manufacturing process, which is in progress now.

Design and manufacture of the ITER Vacuum Vessel

The main functions of the ITER Vacuum Vessel (VV) are to provide the necessary vacuum for plasma operation, act as first nuclear confinement barrier and remove nuclear heating. The design of the VV has been reviewed in the past two years due to more advanced analyses, design modifications required by the interfacing components and R&D. Following the signature of four Procurement Arrangement (PAs), the manufacturing design of the VV sectors, ports and In-Wall Shielding (IWS) is being finalized and the fabrication of the VV sectors has been started in 2012.

ITER’s Tokamak Cooling Water System and the Use of ASME Codes to Comply with French Regulations for Nuclear Pressure Equipment

Abstract During inductive plasma operation of ITER, fusion power will reach 500 MW with an energy multiplication factor of 10. The heat will be transferred by the Tokamak Cooling Water System (TCWS) to the environment using the secondary cooling system. Plasma operations are inherently safe even under the most severe postulated accident condition - a large, in-vessel break that results in a loss-of-coolant accident. A functioning cooling water system is not required to ensure safe shutdown. Even though ITER is inherently safe, TCWS equipment (e.g., heat exchangers, piping, pressurizers) are classified as safety important components. This is because the water is predicted to contain low-levels of radionuclides (e.g., activated corrosion products, tritium) with activity levels high enough to require the design of components to be in accordance with French regulations for nuclear pressure equipment, i.e., the French Order dated 12 December 2005 (ESPN). ESPN has extended the practical application of the methodology established by the Pressure Equipment Directive (97/23/EC) to nuclear pressure equipment, under French Decree 99-1046 dated 13 December 1999, and Order dated 21 December 1999 (ESP). ASME codes and supplementary analyses (e.g., Failure Modes and Effects Analysis) will be used to demonstrate that the TCWS equipment meets these essential safety requirements. TCWS is being designed to provide not only cooling, with a capacity of approximately 1 GW energy removal, but also elevated temperature baking of first-wall/blanket, vacuum vessel, and divertor. Additional TCWS functions include chemical control of water, draining and drying for maintenance, and facilitation of leak detection/localization. The TCWS interfaces with the majority of ITER systems, including the secondary cooling system. U.S. ITER is responsible for design, engineering, and procurement of the TCWS with industry support from an Engineering Services Organization (ESO) (AREVA Federal Services, with support from Northrop Grumman, and OneCIS). ITER–International Organization (ITER-IO) is responsible for design oversight and equipment installation in Cadarache, France. TCWS equipment will be fabricated using ASME design codes with quality assurance and oversight by an Agreed Notified Body (approved by the French regulator) that will ensure regulatory compliance. This paper describes the TCWS design and how U.S. ITER and fabricators will use ASME codes to comply with EU Directives and French Orders and Decrees.

Classification of ITER Tokamak Cooling Water System in Accordance with French Regulations Concerning Pressure and Nuclear Pressure Equipment

Abstract ITER is a joint international fusion facility to demonstrate the scientific and technological feasibility of fusion power for future commercial electric power facilities. ITER is being designed and constructed in France with support from seven domestic agencies. In accordance with the Article 14 of the ITER Agreement, ITER shall observe French Regulations. Among various existing regulatory documents the French Decree 99-1046 concerning pressure equipment and the French Order dated 12th December 2005 concerning nuclear pressure equipment formulate the requirements for design, manufacture and operation of the pressure and nuclear pressure equipment. The ITER Tokamak Cooling Water System (TCWS) is comprised of 4 primary heat transfer systems and their supporting systems. TCWS provides the cooling water to client systems for heat removal during plasma operations and provides the primary confinement for the radioactive substances entrained in the cooling water. The main sources of radioactive substances include Tritium, Activated Corrosion Products (ACP), 14C isotope, 16N and 17N isotope. The concentration of these radioactive substances is a key parameter for the classification of TCWS equipment in accordance with French regulations. The paper will describe the process for classifying TCWS pressure equipment in accordance with French Regulations.