ITER Design Review Research Articles

Development of the ITER baseline inductive scenario

Sustainment of Q ∼ 10 operation with a fusion power of ∼500 MW for several hundred seconds is a key mission goal of the ITER Project. Past calculations and simulations predict that these conditions can be produced in high-confinement mode operation (H-mode) at 15 MA relying on only inductive current drive. Earlier development of 15 MA baseline inductive plasma scenarios provided a focal point for the ITER Design Review conducted in 2007–2008. In the intervening period, detailed predictive simulations, supported by experimental demonstrations in existing tokamaks, allow us to assemble an end-to-end specification of this scenario consistent with the final design of the ITER device. Simulations have encompassed plasma initiation, current ramp-up, plasma burn and current ramp-down, and have included density profiles and thermal transport models producing temperature profiles consistent with edge pedestal conditions present in current fusion experiments. These quasi-stationary conditions are maintained due to the presence of edge-localized modes that limit the edge pressure. High temperatures and densities in the pedestal region produce significant edge bootstrap current that must be considered in modelling of feedback control of shape and vertical stability. In this paper we present new results of transport simulations fully consistent with the final ITER design that remain within allowed limits for the coil system and power supplies. These self-consistent simulations increase our confidence in meeting the challenges of the ITER program.

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.

Equilibrium and vertical-instability considerations for vertical strike-point shifts on the ITER divertor targets

The study of operation with raised strike points on the first ITER divertor target plates is motivated by the need to gain experience with operation with strike points on tungsten (W) surfaces during the non-active phases (in the case of an initial carbon fibre composite (CFC)/W divertor); or (if ITER begins with a full-W divertor), to gain experience with plasma control and transients while operating with raised strike points to avoid damaging the baseline strike regions in preparation for the nuclear phase, and to provide a means for operation should damage occur in the baseline strike zone. For operation with raised strike points, we use the Corsica code to investigate the range of possible H- and L-mode equilibria, with emphasis on the maximum plasma current, achievable shapes, etc. With raised strike points the maximum achievable plasma current is close to 14 MA. The operating space (βp − li) for raised strike points has been studied. The size of the βp − li operating space shrinks (compared to using standard strike-point positions) at 14 MA. For 12 MA, however, the operating space is not affected when using raised strike points. For equilibria with elevated strike points (at roughly the CFC/W transitions, following the 2007 ITER Design Review) the vertical-instability growth-rates at high plasma current (14 MA) are somewhat high but are within the 20 s−1 which studies indicate are controllable in ITER. At lower currents (12 MA) in H-mode, the vertical-instability growth rates stay below 10.0 s−1 for most of βp − li space. At 12 MA in H-mode, multiple equilibria which meet our constraints have been found in overlapping regions of the βp − li operating space.

The design of the ITER first wall panels

The ITER blanket is in the final stage of design completion. The issues raised during the 2007 ITER design review about the first wall (FW) heat loads and remote handling strategy have been addressed, while integrating the recently confirmed in-vessel coils. This paper focuses on the FW design, which is nearing completion. Key design justifications are presented, followed by a summary of the current status of the manufacturing plan and R&D activities.

Design evolution and integration of the ITER in-vessel components

The ITER in-vessel components have experienced a major redesign since the ITER Design Review of 2007. A set of in-vessel vertical stabilization (VS) coils and a set of in-vessel Edge Localized Mode (ELM) control coils have been implemented. The blanket system has been redesigned to include first wall (FW) shaping, to upgrade the FW heat removal capability and to allow for an “in situ” replacement. The blanket manifold system has been redesigned to improve leak detection and localisation. The introduction of a new set of in-vessel coils and the design evolution of the blanket system while the ITER project was entering the procurement phase have proven to be a major engineering challenge. This paper describes the status of the redesign of the in-vessel components and the associated integration issues.

Overview of the design and R&D of the ITER blanket system

Abstract The ITER blanket design has substantially evolved since the ITER design review of 2007. Two major incentives for the design changes have been the need to account for large plasma heat fluxes to the First Wall (FW) and the need for acceptable maintenance of FW panels. In parallel to the design effort, a focused R&D program is being carried out including manufacturing and testing of semi-prototypes for the FW panels, and of full-scale prototypes for the shield blocks. This paper summarizes the status of the ITER blanket system design including the accommodation of interfaces, and describes some of the key R&D activities in support of the design with the goal of starting procurement in the first half of 2013.

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.

Analysis of ITER upper port plug remote handling maintenance scenarios

► Remote Handling Study Centre: providing RH compatibility analysis. ► Simulation: virtual reality including kinematics and realtime physics simulator. ► Applied on analysis of RH compatibility of Upper Launcher component replacement. ► Resulting in lowered maintenance procedure time and lessons learned. The ITER tokamak has a modular design, with port plugs, blanket modules and divertor cassettes. This set-up allows for maintenance of diagnostics, heating systems and first wall elements. The maintenance can be done in situ, or in the Hot Cell. Safe and effective remote handling (RH) will be ensured by the RH requirements and standards. Compliance is verified through remote handling compatibility assessments at the ITER Design Review milestones. The Remote Handling Study Centre at FOM Institute DIFFER is created to study ITER RH maintenance processes at different levels of complexity, from relatively simple situational awareness checks using snap-shots in the CAD system, time studies using virtual reality (VR) animations, to extensive operational sequence validation with multiple operators in real-time. The multi-operator facility mimics an RH work-cell as presently foreseen in the ITER RH control room. Novel VR technology is used to create a realistic setting in which a team of RH operators can interact with virtual ITER environments. A physics engine is used to emulate real-time contact interaction as to provide realistic haptic feed-back. Complex interactions between the RH operators and the control room system software are tested. RH task performance is quantified and operational resource usage estimated. The article provides a description and lessons learned from a recent study on replacement of the Steering Mirror Assembly on the ECRH (Electron Cyclotron Resonance Heating) Upper Launcher port plug.

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.

Overview of the ITER EC H&CD system and its capabilities

The Electron Cyclotron (EC) system for the ITER tokamak is designed to inject ≥20 MW RF power into the plasma for Heating and Current Drive (H&CD) applications. The EC system consists of up to 26 gyrotrons (between 1 and 2 MW each), the associated power supplies, 24 transmission lines and 5 launchers. The EC system has a diverse range of applications including central heating and current drive, current profile tailoring and control of plasma magneto-hydrodynamic (MHD) instabilities such as the sawtooth and neoclassical tearing modes (NTMs). This diverse range of applications requires the launchers to be capable of depositing the EC power across nearly the entire plasma cross section. This is achieved by two types of antennas: an equatorial port launcher (capable of injecting up to 20 MW from the plasma axis to mid-radius) and four upper port launchers providing access from inside of mid radius to near the plasma edge. The equatorial launcher design is optimized for central heating, current drive and profile tailoring, while the upper launcher should provide a very focused and peaked current density profile to control the plasma instabilities. The overall EC system has been modified during the past 3 years taking into account the issues identified in the ITER design review from 2007 and 2008 as well as integrating new technologies. This paper will review the principal objectives of the EC system, modifications made during the past 2 years and how the design is compliant with the principal objectives.

Physics basis and design of the ITER plasma-facing components

In ITER, as in any tokamak, the first wall and divertor plasma-facing components (PFC) must provide adequate protection of in-vessel structures, sufficient heat exhaust capability and be compatible with the requirements of plasma purity. These functions take on new significance in ITER, which will combine long pulse, high power operation with severe restrictions on permitted core impurity concentrations and which, in addition, will produce transient energy loads on a scale unattainable in today’s devices. The current ITER PFC design has now reached a rather mature stage following the 2007 ITER Design Review. This paper presents the key elements of the design, reviews the physics drivers, essentially thermal load specifications, which have defined the concept and discusses a selection of material and design issues.

R&D of atmosphere detritiation system for ITER in JAEA

In order to establish an effective ITER atmosphere detritiation system (DS), JAEA (Japan Atomic Energy Agency) has investigated the performance and the durability of the system at various incident/accident conditions in support of finalizing the DS conceptual design through the ITER design review. The current DS as the Safety Important Component (SIC) has been discussed and mainly consists of catalytic reactors, wet scrubber columns (SCs), and blowers. The functional failure of the DS designed with SC was evaluated using a database of failure experiences of various valves, controllers, and components. Regarding the tritium release into the biggest confinement sector of the Tokamak gallery, this design is improved by more than two orders of magnitude compared to the original DS designed with molecular sieves (MSs) dryer beds in the 2001 design report. This improvement is achieved mainly by the reduction of frequency of valve operation, like MS dryers requiring periodical regeneration, and by the standardized module arrangement of the DS with SC.

Status and physics basis of the ITER divertor

The ITER divertor design is the culmination of years of physics and engineering effort, building confidence that this critical component will satisfy the requirements and meet the challenge of burning plasma operation. With 54 cassette assemblies, each weighing ∼9 tonnes, nearly 3900 actively cooled high heat flux elements rated to steady-state surface power flux densities of 10 MW m−2 and a total of ∼60 000 carbon fibre composite monoblocks and ∼260 000 tungsten monoblocks/flat tiles, the ITER divertor will be the largest and most advanced of its kind ever constructed. Both the ITER Design Review and subsequent follow-up activities have led to a number of modifications to the device, including the divertor design, significantly improving ITER's operational flexibility. This paper outlines the salient features of the final divertor design, with emphasis on the physics rationale that has defined the design choices and on the performance of the resulting configuration.

22nd IAEA Fusion Energy Conference: summary of contributions on Fusion Technology and ITER Activities

The contributions to the 22nd IAEA Fusion Energy Conference (FEC) in the categories of Fusion Technology (FT), ITER Activities (IT) and Safety and Economic Studies (SE) are reviewed. In the FT category, 68 papers were submitted, along with 57 papers submitted through the ITER Organisation in the IT category. Finally two papers were submitted in the SE category. The assembled body of work gave a good overview of the worldwide effort in fusion technology and particularly the prolific activity surrounding the ITER Design Review and the major progress with the ITER technology programme.

Integrated modelling of ITER reference scenarios

The ITER Scenario Modelling Working Group (ISM WG) is organized within the European Task Force on Integrated Tokamak Modelling (ITM-TF). The main responsibility of the WG is to advance a pan-European approach to integrated predictive modelling of ITER plasmas with the emphasis on urgent issues, identified during the ITER Design Review. Three major topics are discussed, which are considered as urgent and where the WG has the best possible expertize. These are modelling of current profile control, modelling of density control and impurity control in ITER (the last two topics involve modelling of both core and SOL plasma). Different methods of heating and current drive are tested as controllers for the current profile tailoring during the current ramp-up in ITER. These include Ohmic, NBI, ECRH and LHCD methods. Simulation results elucidate the available operational margins and rank different methods according to their ability to meet different requirements. A range of ‘ITER-relevant’ plasmas from existing tokamaks were modelled. Simulations confirmed that the theory-based transport model, GLF23, reproduces the density profile reasonably well and can be used to assess ITER profiles with both pellet injection and gas puffing. In addition, simulations of the SOL plasma were launched using both H-mode and L-mode models for perpendicular transport within the edge barrier and in the SOL. Finally, an integrated approach was also used for the predictive modelling of impurity accumulation in ITER. This includes helium ash, extrinsic impurities (such as argon) and impurities coming from the wall (including tungsten). The relative importance of anomalous and neo-classical pinch contributions towards impurity penetration through the edge transport barrier and further accumulation in the core was assessed.

Principal physics developments evaluated in the ITER design review

As part of the ITER Design Review and in response to the issues identified by the Science and Technology Advisory Committee, the ITER physics requirements were reviewed and as appropriate updated. The focus of this paper will be on recent work affecting the ITER design with special emphasis on topics affecting near-term procurement arrangements. This paper will describe results on: design sensitivity studies, poloidal field coil requirements, vertical stability, effect of toroidal field ripple on thermal confinement, material choice and heat load requirements for plasma-facing components, edge localized modes control, resistive wall mode control, disruptions and disruption mitigation.

ITER vertical stabilization system

As part of the ITER design review, a reassessment of the specifications underlying the design of the vertical stabilization system (VS) was performed. Recent results from experiments, aimed at the evaluation of the feasibility of the ITER reference scenarios, have raised several concerns regarding mostly the ramp-up and ramp-down phases of the pulse. The main issue is the value of the internal inductance li which may reach values outside the range 0.7–1, considered as reference for the ITER control system design. Similar concerns apply to the low current L-mode plasmas, needed to the exploitation of the machine towards the development of the 15 MA pulse. The performance of the reference vertical stabilization system, under the revised conditions may be marginal, in particular if the effect of plasma generated noise on the velocity measurement is considered. A reliable and robust VS is mandatory to guarantee the operation of ITER at the reference elongation and plasma current values. To avoid de-scoping of the machine mission, several solutions have been proposed to improve the VS performances, ranging from an upgrade of the maximum voltage available to the present external coils system, to the introduction of in-vessel passive and/or active conductors. The paper presents an overview of the modelling and experimental effort aimed at the assessment of the baseline ITER VS and analyses the proposed solutions to improve the system performance.

Modified blanket cooling manifold system for ITER

The modular ITER shielding blanket is the innermost part of the reactor directly exposed to the plasma. Its high thermal and nuclear loaded blanket modules (BMs) are water-cooled components. The cooling water is supplied to the BMs by a set of stiff inlet and outlet manifolds mechanically attached to the inner wall of vacuum vessel (VV). The ITER reference design of the blanket cooling manifold (BCM) presents some concerns namely: huge reaction forces at the end supports on the VV, high operating stresses at certain locations on the manifolds the difficulty of localizing possible water leakages and the inability to repair the components remotely in a case of manifold damage. To overcome these problems, a proposal for a new manifold concept has been developed within the European Fusion Development Agreement – Central Support Unit in Garching in the frame of the ITER design review. The original stiff welded structure has been replaced by a bundle of elastic single seamless pipes supplying the cooling water to each module. Based on FEM and CFD calculations the shape and diameter for the pipes were chosen to cope with thermal differential expansion as well as with electro-magnetic (EM) loads and pressure drop requirements. An attachment system was developed to enable precise positioning and fixing of the manifold segments onto the inner wall of the VV. Additionally a remote handling (RH) procedure for the removal and installation sequences was created for the case of manifold damage during operation. This procedure is based on the currently planned RH equipment such as the maintenance cask and the in-vessel transporter (IVT).

Low pressure and high power rf sources for negative hydrogen ions for fusion applications (ITER neutral beam injection) (invited)

The international fusion experiment ITER requires for the plasma heating and current drive a neutral beam injection system based on negative hydrogen ion sources at 0.3 Pa. The ion source must deliver a current of 40 A D(-) for up to 1 h with an accelerated current density of 200 Am/(2) and a ratio of coextracted electrons to ions below 1. The extraction area is 0.2 m(2) from an aperture array with an envelope of 1.5 x 0.6 m(2). A high power rf-driven negative ion source has been successfully developed at the Max-Planck Institute for Plasma Physics (IPP) at three test facilities in parallel. Current densities of 330 and 230 Am/(2) have been achieved for hydrogen and deuterium, respectively, at a pressure of 0.3 Pa and an electron/ion ratio below 1 for a small extraction area (0.007 m(2)) and short pulses (<4 s). In the long pulse experiment, equipped with an extraction area of 0.02 m(2), the pulse length has been extended to 3600 s. A large rf source, with the width and half the height of the ITER source but without extraction system, is intended to demonstrate the size scaling and plasma homogeneity of rf ion sources. The source operates routinely now. First results on plasma homogeneity obtained from optical emission spectroscopy and Langmuir probes are very promising. Based on the success of the IPP development program, the high power rf-driven negative ion source has been chosen recently for the ITER beam systems in the ITER design review process.