ITER Lifetime Research Articles

Design and integration of lower ports for ITER diagnostic systems

All around the ITER vacuum vessel, forty-four ports will provide access to the vacuum vessel for remote handling operations, diagnostic systems, heating, and vacuum systems: 18 upper ports, 17 equatorial ports, and 9 lower ports. Among the lower ports, three of them will be used for the remote handling installation of the ITER divertor. Once the divertor is in place, these ports will host various diagnostic systems mounted in the so-called diagnostic racks. The diagnostic racks must allow the support and cooling of the diagnostics, extraction of the required diagnostic signals, and providing access and maintainability while minimizing the leakage of radiation toward the back of the port where the humans are allowed to enter. A fully integrated inner rack, carrying the near plasma diagnostic components, will be an stainless steel structure, 4.2m long, with a maximum weight of 10t. This structure brings water for cooling and baking at maximum temperature of 240°C and provides connection with gas, vacuum and electric services. Additional racks (placed away from plasma and not requiring cooling) may be required for the support of some particular diagnostic components. The diagnostics racks and its associated ex vessel structures, which are in its conceptual design phase, are being designed to survive the lifetime of ITER of 20 years. This paper presents the current state of development including interfaces, diagnostic integration, operation and maintenance, shielding requirements, remote handling, loads cases and discussion of the main challenges coming from the severe environment and engineering requirements.

Final design of the generic equatorial port plug structure for ITER diagnostic systems

The Diagnostic Generic Equatorial Port Plug (GEPP) is designed to be common to all equatorial port-based diagnostic systems. It is designed to survive throughout the lifetime of ITER for 20 years, 30,000 discharges, and 3000 disruptions. The EPP structure dimensions (without Diagnostic First Walls and Diagnostic Shield Modules) are L2.9×W1.9×H2.4m3. The length of the fully integrated EPP is 3174mm. The weight of the EPP structure is about 15t, whereas the total weight of the integrated EPP may be up to 45t. The EPP structure provides a flexible platform for a variety of diagnostics. The Diagnostic Shield Module assemblies, or drawers, allow a modular approach with respect to diagnostic integration and maintenance. In the nuclear phase of ITER operations, they will be remotely inserted into the EPP structure in the Hot Cell Facility. The port plug structure must also contribute to the nuclear shielding, or plugging, of the port and further contain circulated water to allow cooling during operation and heating during bake-out. The Final Design of the GEPP has been successfully passed in late 2013 and is now heading toward manufacturing. The final design of the GEPP includes interfaces, manufacturing, R&D, operation and maintenance, load cases and analysis of failure modes.

Tritium absorption and desorption in ITER relevant materials: comparative study of tungsten dust and massive samples

Tritium adsorption and desorption from well characterized tungsten dust are presented. The dust used are of different types prepared by planetary milling and by aggregation technique in plasma. For the milled powder, the surface specific area (SSA) is 15.5m2/g. The particles are poly-disperse with a maximum size of 200nm for the milled powder and 100nm for the aggregation one. Prior to tritiation the particles are carefully de-oxidized. Both samples are experiencing a high tritium inventory from 5GBq/g to 35GBq/g. From comparison with massive samples and considering that tritium inventory increases with SSA, it is shown that surface effects are predominant in the tritium trapping process. Extrapolation to the ITER environment is undertaken with the help of a Macroscopic Rate Equation model. It is shown that, during the life time of ITER, these particles can exceed rapidly 1GBq/g.

Status and verification strategy for ITER neutronics

The paper summarizes the current status of neutronics at ITER and a first set of proposals for experimental programmes to be conducted in the early operational life-time of ITER are described for the more crucial areas. These include a TF coils heating benchmark, a streaming benchmark and streaming measurements by activation on ITER itself. Also on ITER the measurement of activated water from triton burn-up should be planned and performed. This will require the measurement of triton burn-up in DD phase. Measurements of neutron flux in the tokamak building during DD operations should also be carried out. The use of JET for verification of shut down dose rate estimates is desirable. Other facilities to examine the production and behaviour of activated corrosion products and the shielding properties of concretes to high energy (6MeV) gamma-rays are recommended.

ITERDB—The Data Archiving System for ITER

For ITER, acquiring, managing and archiving its data is an essential task. ITER is foreseen to produce up to one terabyte of data per pulse and several petabytes of data per year. All the produced data needs to be stored and managed. The stored data is expected to serve the data access needs of ITER researchers located both on the ITER premises as well as worldwide during ITER's lifetime and beyond.ITERDB is a data management system being designed for centralized ITER data archival and data access. It is designed to manage and serve both unprocessed and processed data from the ITER plant systems and data analysis workflows.In this paper, we report the ITER Data Archiving System software requirements and priorities that have been identified by working with ITER staff and a large number of stakeholders. We will describe the design challenges and the proposed solutions. We will also present the current state of the ITERDB software architecture design.

Software Quality Assurance activities of ITER CODAC

Software as an integral part of the plant system I&C is crucial in the manufacturing and integrated operation of ITER plant systems. Software Quality Assurance is necessary to ensure the development and maintenance of consistently high quality I&C software throughout the lifetime of ITER. CODAC decided to follow IEEE 12207-2008 software lifecycle processes for Software Engineering and Software Quality Assurance. Software Development Plan, Software Configuration Management Plan and Software Verification and Validation Plan are the mainstay of Software Quality Assurance which is documented in the Software Quality Assurance Plan. This paper describes the Software Quality Assurance (SQA) activities performed by CODAC. The SQA includes development and maintenance of above plans, processes and resources. With the help of Verification and Validation Teams they gather evidence of process conformance and product conformance, and record process data for quality audits and perform process improvements.

Nuclear analysis of the ITER full-tungsten divertor

This paper presents the nuclear analysis performed for the ITER full-tungsten divertor using the MCNP-5 Monte Carlo Code in a 3-D geometry. A detailed model of the geometry of the divertor based on the last design specifications has been integrated into the latest ITER MCNP model. Nuclear heating, damage and helium production have been calculated. The presented results are consistent with recent analysis performed with ATTILA code and with the previous ones with MCNP. The shielding capabilities of the last design are reduced in comparison with the design of 2004 but a negligible impact is expected. The reweldability of radial pipes of the cassette body (CB) is not a concern even assuming irradiation during the whole ITER lifetime. The reweldability of pipes for the refurbishment of the CB with new plasma facing components depends on maintenance scenario. It is not expected that last 2012 divertor design leads to an appreciable increase of the nuclear loads on vacuum vessel and toroidal field coils, since the relative contribution of the divertor region remains low.

The fundamental role of fluid dynamic analyses in the design of the solid EU Test Blanket Module

In the frame of the activities of the European Test Blanket Module Consortium of Associates, the Helium Cooled Pebble Bed Test Blanket Module (HCPB TBM), the so-called solid TBM, is developed in Karlsruhe Institute of Technology (KIT). In the EU experimental strategy, a series of 4 different HCPB TBMs will be connected to the dedicated equatorial port n.16 during the ITER lifetime. The ITER TBM program has to provide DEMO relevant experimental data for the main functions of the blanket modules of a future fusion reactor.The preliminary thermo-mechanical design assessment of the TBM box (based on transient, steady state and accidental analyses) has been presented. All along the design assessment phase the fluid dynamic analyses play a fundamental role for the TBM sub-components, the Breeder Units (BUs) and the manifolds (MF) stages. This paper highlights the methodology implemented for the Computational Fluid Dynamic (CFD) analyses in the TBM design life cycle, and presents the results and the impact on the overall performance evaluation of the HCPB TBM. The following models are presented in detail: the CFD model of the TBM First Wall and its application to a reduced scale First Wall, the 3D CFD model of the BUs, and the thermo fluid dynamic modelling of the manifold systems.

Optimization of the availability of the core CXRS diagnostics for ITER

New optical configurations for the ITER core CXRS system offer the possibility of longer ducts between the first mirror and the plasma. This has led to a renewed optimization of the availability, using a simple model of the degradation of the first mirror that starts with the conditions of (a) the required measurement performance and (b) the geometry of the port plug. It is found that for a fully passive system the design should strive for the longest duct length possible. Given known data, this will result in a diagnostic lifetime still substantially shorter than ITER lifetime. When an option of cleaning the first mirror is introduced (assuming this is a feasible option) the optimum is less straightforward, because the lifetime of the second mirror then also becomes important. The optimum then depends on the ratio between the cleaning interval and the ITER lifetime. Options are presented for various sets of assumptions. Finally practical limitations of supporting subsystems (cleaning system, shutter, calibration system) may influence the final design. Examples of such limitations with their impact are presented.

ITER In-Cryostat inspection and repair feasibility studies

The ITER In-Cryostat maintenance study is an important precondition to guarantee the operation over the ITER lifetime. The ITER operation is subdivided mainly into two phases: 1. Hydrogen phase (non-nuclear operation phase). 2. Deuterium/Tritium phase (nuclear DT phase). The commissioning phase includes the initial phase of assembly. Within the first phase the ITER components will be tested; afterwards they will go into operation. The In-Cryostat maintenance shall facilitate all operations that could be required by In-Cryostat systems and the Cryostat itself. In cases of failures or unlikely events (e.g. earthquakes) it is necessary to provide man and tool access to In-Cryostat components. Overall functions which have to be implemented are: • Inspection of components including leak localization (helium, water, air). • Repair and replacement of component (instrumentation, parts or complete components). • Regulatory inspections. It is presumed that most of component failure would occur at the beginning of the operational phase. This failure rate is expected to be very unlikely when ITER is being operating during the nuclear phase. For maintenance activities it is assumed that: • The intervention frequency on each component is limited during its lifetime (e.g. inspections/repair during global shutdown). • Most of these interventions will be required during the inactive phase. According to ALARA (As Low as Reasonable Achievable) rules maintenance activities will be planned in order to minimize the required human interventions during the active phase. Different tools have to be designed to perform the maintenance actions. As there are quiet all heavy components to be handled and removed, humans cannot perform the work without semi hands-on tools. The required permanent fixtures and tools are considered and pre-designed.

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.

Assessment of the ITER high-frequency magnetic diagnostic set

The ITER high-frequency (HF) magnetic diagnostic system has to provide essential measurements of MHD instabilities with | δB MEAS/ B POL| ∼ 10 −4 (∼1 G) for frequencies up to 2 MHz to resolve toroidal mode numbers ( n) in the range | n| = 10 to | n| = 50. A review of the measurement requirements for HF MHD instabilities in ITER was initiated during the TW4 work-program and led to significant interest for physics and real-time control issues in measuring modes with | δB MEAS| as low as ∼10 −3 G at the position of the sensors, with | n| ≤ 30 and poloidal mode numbers | m| ∼ 2| n| up to | n| ∼ 15, for a frequency range extending up to ∼500 kHz. We have examined the ability of the current ITER design for the individual sensors and the diagnostic system as a whole to meet these needs, and have explored what adjustments to the design (of the individual sensors and/or of the system as a whole) or to the requirements would be needed to meet them when considering different hypothesis for the financial costs and risk management over the ITER life-time. First, we find that the proposed diagnostic layout, with 168 sensors in total, does not meet the more stringent measurement requirements and risk management criteria: these can only be met by a revision of the design requiring 350–500 sensors, depending on different costing and risk management options. Second, we find that the current design for the ITER HF Mirnov-type pick-up coil could be usefully revised.

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.

The European Breeding Blanket Test Facility: An integrated device to test European helium cooled TBMs in view of ITER

The activities of testing and qualification of the two European Test Blanket Modules (TBMs) to be installed in ITER as well as their main ancillary circuits are of outstanding importance in view of the experimental activities planned during ITER lifetime. Having in mind this purpose the EBBTF has been designed and installed at ENEA Brasimone Research Centre in order to meet the important requirements coming directly from the scientific communities to tests and qualify blanket components and technologies in view of ITER and DEMO reactors. ENEA is responsible for the activities of design, construction and installation of the system at ENEA Brasimone and for all of them different national companies have been involved. Essentially, EBBTF consists of the upgraded HeFus3 helium loop coupled with a new lead lithium loop named IELLLO ( Integrated European Lead Lithium LOop). The two loops working together will be able to test the HCLL and HCPB mock-up, up to 1:1 scale. The device and its original design solution are described in this paper.

Review of design principles for ITER VV remote inspection in magnetic field

Because ITER magnet system has a limited number of mechanical and thermal stress cycles, shut down number of the toroidal field is limited during lifetime of ITER. Any inspection device able to withstand the toroidal field between two plasma shots will enhance the inspection frequency capacity of ITER during operation phase. In addition to the high magnetic field the system should also cope with high temperature, ultra-high vacuum and high radiation, in order to keep the reactor availability high. Radiation, ultra-high vacuum and temperature constraints already addressed by on going R&D activities within Europe—considering the required level of radiation is to date the highest encountered in remote handling, and that facing all these constraints at once is an additional issue to overcome. Whereas, operating remote handling systems in high magnetic field is quite new field of investigation. This paper aims to be a guideline for future designers to help them choose among options the adequate solution for an ITER relevant inspection device. It provides the designer an objective view of the different effects that stem from technical choices and help them deciding whether a technology is relevant or not depending on the task's requirements. We have selected a set of technologies and products available for structural design, actuation, sensing and data transmission in order to design inspection remote handling equipment for ITER in the given constraints. These different solutions are commented with specific considerations and directions to have them fit in the specifications. Different design strategies to cope with magnetic field are then discussed, which imply either insensitive design or using the magnetic field as a potential energy source and as a positioning help. This analysis is the first result of one of the projects in the PREFIT partnership, part of the European Fusion Training Scheme.

Activation and afterheat analyses for the HCPB test blanket module in ITER

To provide a sound data basis for the safety analyses of the HCPB TBM system in ITER, the afterheat and activity inventories were assessed making use of a code system that allows performing 3D activation calculations by linking the Monte Carlo transport code MCNP and the fusion inventory code FISPACT through an appropriate interface. A suitable MCNP model of a 20° ITER torus sector with an integrated TBM of the HCPB PI (plant integration) type in the horizontal test blanket port was developed and adapted to the requirements for coupled 3D neutron transport and activation calculations. Two different irradiation scenarios were considered in the coupled 3D neutron transport and activation calculations. The first one is representative for the TBM irradiation in ITER with a total of 9000 neutron pulses over a 3 (calendar) years period. The second (conservative) irradiation scenario assumes an extended irradiation time over the full anticipated lifetime of ITER. The radioactivity inventories, the afterheat and the contact gamma dose were calculated as function of the decay time. Data were processed for the total activity, afterheat and contact dose rates of the TBM, its constituting components and materials.

Dust control in tokamak environment

During ITER lifetime, dusts and flakes will be produced due to the interaction of plasmas with the in-vessel materials or due to maintenance. They will be made of carbon, beryllium and tungsten and will be activated, tritiated and chemically reactive and toxic. Safety limits have been set in order to reduce dust hazards. Thus, dust diagnostics and removal methods need to be developed for ITER considering the constraints linked to magnetic field, radiation, vacuum and temperature. This paper reviews potential diagnostics to monitor the dust content based on erosion, deposition or dust in suspension measurements using optical, sampling or gravimetric systems.

The ECH front steering launcher for the ITER upper port

The purpose of the ITER electron cyclotron resonance heating (ECRH) upper port antenna will be to drive current locally inside the island which forms on the q = 3/2 or 2 rational magnetic flux surfaces in order to stabilize neoclassical tearing modes (NTMs). The launcher should be capable of steering the focused beam deposition location across the resonant flux surface over the range in which the q = 3/2 and 2 surfaces are found, for the various plasma equilibria susceptible to the onset of NTMs. ITER's present reference design uses a front steering (FS) concept, with the moveable mirror close to the plasma. Two separate mirrors are used to decouple the focussing and steering aspects resulting in an optimized folded beam path. The FS launcher is capable of steering eight 2 MW beams via two sets of steering mirrors out of each of the four allocated upper port plugs. Among the relevant mechanical design aspects are the systematic reliance on frictionless and backlash free mechanical movements based on the compliant deformation of structural components to avoid the invessel tribological difficulties [J.-D. Landis, et al., Design of the critical components in the ITER ECH upper launcher steering mechanism, this conference]. An inert gas pressure controlled bellows system provides accurate angular positioning of the steering mirror. The antenna is designed fail-safe to provide full bandwidth beam steering capabilities during the twenty years lifetime of ITER. Details of the FS launcher design relating to mm-wave, thermohydraulic, electromagnetic, neutron shielding and structural aspects and possible options regarding the integration of launcher components into the port plug are discussed.

Design of the critical components in the ITER ECH upper launcher steering mechanism

The purpose of the ITER electron cyclotron resonance heating (ECRH) antennae located in the upper port launcher will be to stabilize the neoclassical tearing modes (NTM) by driving currents locally inside either the q = 3 / 2 or 2 island. The ITER reference design uses a front steering (FS) concept with the moveable mirror close to the plasma. The FS launcher is capable of steering eight 2 MW beams via two sets of steering mirrors [R. Chavan, M. Henderson, R. Bertizzolo, J.-D. Landis, F. Sanchez, H. Shidara, The ECH front steering launcher for the ITER upper port, this conference. [8]]. In the adverse invessel operating conditions, reliable operation is required to guarantee the availability of the mirror steering mechanism during the 20-year lifetime of ITER. The dynamic performance and steering accuracy of the mirror system are essential for the localized beam power deposition. In order to increase the steering reliability and accuracy, traditional bearing and push–pull rods are avoided, which tend to introduce backlash or grip. The proposed frictionless and backlash free mechanism design uses elastically compliant structural components to guide and drive the rotating mirror. The actuator system is based on a pressure controlled helium filled bellows array working against a set of preloaded compressive springs. The radiation and temperature sensitive pressure control servovalves are located in the port duct behind the closure plate. Resistance to cyclic fatigue is the critical design requirement for compliant systems. In order to achieve the high reliability, four critical components are identified and described in detail. The bellows, the flexure pivots, the return springs and the spiral cooling tubes are the core elements of the mechanism allowing the controlled rotation of the mirror. Prototypes of the critical components are being tested. Corresponding analytical and cyclic fatigue numerical models of the compliant components are to be validated by fatigue bench tests. The detailed design and the principal tests of the critical components used in the FS mechanism will be presented.

Investigation of stability of ITER candidate Hall sensors under neutron irradiation

Hall sensors offer an attractive true non-inductive method of magnetic field measurements for fusion reactors. Their use for steady state magnetic diagnostics of ITER is presently limited by their questionable radiation and thermal stability. Issues of stable and reliable operation in ITER like radiation and thermal environment are addressed by the contribution. Recently, novel Hall sensors, compatible with temperatures up to 200°C, were developed and their radiation stability was tested at LVR-15 experimental fission reactor. Overview of the experimental set-up on LVR-15 reactor is given. Degradation of the sensor’s sensitivity by several tens of percents was observed after neutron irradiation by the total neutron fluence of 2 × 1017 n/cm2 in LVR-15. This level of neutron fluence is comparable to that expected to occur over the whole ITER life time for a sensor location just outside the ITER vessel. The in-situ recalibration techniques are expected to handle the observed degree of Hall sensors performance degradation in ITER environment.