Technological review of the HRP manufacturing process R&D activity

ENEA and Ansaldo Nucleare S.p.A. have been deeply involved in the European International Thermonuclear Experimental Reactor (ITER) R&D activities for the manufacturing of high heat flux plasma-facing components (HHFC), and in particular for the inner vertical target (IVT) of the ITER divertor.This component has to be manufactured by using both armour and structural materials whose properties are defined by ITER. Their physical properties prevent the use of standard joining techniques. The reference armour materials are tungsten and carbon/carbon fibre composite (CFC). The cooling pipe is made of copper alloy (CuCrZr-IG).During the last years ENEA and Ansaldo have jointly manufactured several actively cooled monoblock mock-ups and prototypical components of different length, geometry and materials, by using innovative processes: HRP (hot radial pressing) and PBC (pre-brazed casting).The history of the technical issues solved during the R&D phase and the improvements implemented to the assembling tools and equipments are reviewed in the paper together with the testing results.The optimization of the processes started from the successful manufacturing of both W and CFC armoured small scale mockups thermal fatigue tested in the worst ITER operating condition (20 MW/m²) through the achievement of record performances obtained from a monoblock medium scale mockup.On the base of these results ENEA-ANSALDO participated to the European programme for the qualification of the manufacturing technology to be used for the procurement of the ITER divertor IVT, according to the F4E specifications. A divertor inner vertical target prototype (400 mm total length) with three plasma facing component units, was successfully tested at ITER relevant thermal heat fluxes.Now, ANSALDO and ENEA are ready to face the challenge of the ITER inner vertical target production, transferring to an industrial production line the experience gained in the development, optimization and qualification of the PBC and HRP processes.     

Engineering studies for integration of the test blanket module (TBM) systems inside an ITER equatorial port plug

The European test blanket module (EU-TBM), first prototype of the breeding blanket concepts under development for the future DEMO power plant to produce the tritium, will be developed to be tested in three equatorial ports of ITER dedicated to this. The CEA Cadarache under the contract of Association EURATOM/CEA and in close relation with Association EURATOM/HAS works on the integration of the EU-TBM inside ITER tokamak.The installation of the TBM into the vacuum vessel is made with the help of a port plug, constituted with two components: the Shield module and the Port-Plug frame. The Shield module provides the neutron shielding inside the Port-Plug frame, which maintains in cantilever position the TBM and its shield module and closes the vacuum vessel port.This paper will describe the EU-TBM design and integration activities on the cooled shield module and on its interface with the TBM component. A particular attention, in term of thermal and mechanical studies, is dedicated to the design of the shield and test blanket module attachment, and also to the shield design and its internal cooling system.     

The WEST project: PFC shaping solutions investigated for the ITER-like W divertor

Among major issues for PFCs design, the impact of leading edges (exposed surface) which would be directly intersected by particles following magnetic field lines at glancing incident angles in the high heat flux areas is much discussed. This paper presents the key outcome of a thermal analysis performed on different shaping solutions for the ITER-like W monoblocks occurred for the components of the WEST (W Environment for Steady state Tokamak) divertor which could shadow any direct leading edge and to counteract a potential misalignment due to assembly tolerance. The results, in terms of surface temperature rise and wall heat flux into the cooling channel, are discussed for magnetic field lines incident at glancing angles expected in the higher heat flux regions of divertor (i.e. close to the strike point regions) and for perpendicular incident heat flux up to 20 MW/m².     

Thermo-hydraulic analysis of the generic equatorial port plug design

The port-based ITER diagnostic systems are housed primarily in two locations, the equatorial and upper port plugs. The port plug structure provides confinement function, maintains ultra-high vacuum quality and the first confinement barrier for radioactive materials at the ports. The port plug structure design, from the ITER International Organisation (IO), is cooled and heated by pressurized water which flows through a series of gun-drilled water channels and water pipes. The cooling function is required to remove nuclear heating due to radiation during operation of ITER, while the heating function is intended to heat up uniformly the machine during baking condition. The work presented provides coupled thermo-hydraulic analysis and optimization of a Generic Equatorial Port Plug (GEPP) structure cooling and heating system. The optimization performed includes positioning, minimization of number and size of gun drilled channels, complying with the flow and functional requirements during operating and baking conditions.     

Electromagnetic analysis of ITER generic equatorial port plug designs during three plasma current disruption cases

Electromagnetic phenomena due to plasma current disruptions are the cause for the main mechanical operation loads over the ITER equatorial level port plug structures.This paper presents a detailed finite element simulation and analysis of the transient electromagnetic effects of three different plasma current disruption cases over three designs of diagnostic shielding module (DSM) structure. The DSMs are contained into and supported by the generic equatorial port plug (GEPP) analyzed structure.The three plasma disruption cases studied were: major disruption upwards linear decay in 36 ms (MD UP LIN36), vertical displacements events, upwards and downwards linear decay in 36 ms (VDE UP LIN36 and VDE DW LIN36). A detailed analysis for GEPP structure and three DSM-first wall (FW) designs (horizontal and vertical drawers and monoblock) is also presented in order to extract the Eddy current distribution on these devices and thus the resultant electromagnetic forces and moments acting on them.     

Experiments at the T-11M device in support of the tokamak concept with closed Li cycle

The lithium emitter-collector concept, supposed the creation of steady state lithium circulation loop close the first wall and plasma periphery, seems as a solution of tokamak reactor first wall problem. This concept of renewable PFC (plasma facing components) has four characteristically steps: Li emission from PFC into the plasma (emitter), the boundary plasma cooling by non-coronal Li radiation, Li ions capture by collector before their deposit on a tokamak vessel wall and Li return from collector and first wall into emitter again. The main subject of the last T-11M experiments was investigations of the lithium collection by limiters. The lithium behavior in shadow of lithium limiter, based on the lithium filled CPS (capillary porous systems), was studied by the analysis of the witness-samples, and by use a mobile graphite probe. It was showed: (1) lithium deposit on the Li-limiter sides is proportional to the total Li emission from the lithium rail limiter (emitter). (2) Lithium deposit on the ion-drift side of Li-limiter to 3-5 times more as on electron side. (3) The total efficiency of Li collection by T-11M limiters can be 70 ± 20% of lithium integral emission from the lithium emitter during plasma operations while the theoretically limit can be 90%.     

Progresses and challenges in supporting activities toward a license to operate European TBM systems in ITER

Licensing a pressurized nuclear equipment like the European Test Blanket Modules (TBM) Systems and, on the longer term, breeder blankets of a fusion demonstration reactor (DEMO), will require presenting to the Regulator and the Agreed Notified Body, along with design and safety analyses, supporting data like consolidated materials data and design limits, qualified fabrication procedures specifications and validated modeling tools that go often over today's state-of-the-art of nuclear industry. TBM systems feature indeed a newly developed structural material and advanced fabrication processes that were not referenced in any nuclear construction codes before, new type of functional materials, complex structures geometry and many interconnected sub-systems exchanging tritium by permeation or fluid mass transfer. For many years now, Europe has structured its development activities on TBM Systems toward the preparation of licensing. First tangible results are now arising: the EUROFER structural material has been introduced in the RCC-MRx nuclear code, supported by a database of several thousands of test records; TBM box fabrication procedure specifications are under standardization by industry in view of their qualification; a modeling tool for accurate simulation of tritium transport in TBM systems has been developed in view of refining conservative inventory data published in preliminary safety reports and optimizing waste management. Remaining challenges are identified and discussed.     

Preliminary design of a helium cooled molten lithium test blanket module for the ITER test in Korea

Through a consideration of the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He cooled molten lithium (HCML) blanket with ferritic steel (FS) as a structural material in the International Thermonuclear Experimental Reactor (ITER) program. The preliminary design and its performance of KO HCML test blanket module (TBM) are introduced in this paper. It uses He as a coolant at an inlet temperature of 300 °C and an outlet temperature up to 400 °C and Li is used as a tritium breeder by considering its potential advantages. Two layers of graphite are inserted as a reflector in the breeder zone to increase the tritium breeding ratio (TBR) and the shielding performances. A 3-D Monte Carlo analysis is performed with the MCCARD code for the neutronics and the total TBM power is designed to be 0.739 MW at a normal heat flux from the plasma side. From the analysis results with CFX-10 for the thermal-hydraulics, the He cooling path is determined and it shows that the maximum temperature of the first wall does not exceed 550 °C at the structural materials and the coolant velocities are 45 and 11.5 m/s in the first wall and breeding zone, respectively. The obtained temperature data is used in the thermal-mechanical analysis with ANSYS-10. The maximum von Mises equivalent stress of the first wall is 123 MPa and the maximum deformation of it is 3.73 mm, which is lower than the maximum allowable stress.     

Experience and technical issues of liquid lithium application as plasma facing material in tokamaks

The following critical issues of liquid lithium used in tokamak conditions are considered: major physical properties of lithium, physico-chemical aspects of lithium interaction and compatibility with structural materials of fusion reactors. Lithium capillary-porous system (CPS) is considered as advanced plasma facing material for power fusion reactor and its main properties are presented. Review of plasma facing element (PFE) structures based on lithium CPS and tests results in T-11M, T-10 and FTU tokamaks are included. Brief review of projects of lithium limiter of FTU with active system for thermal stabilization and module of lithium divertor for KTM tokamak with liquid metal (Na-K) cooling system based on the lithium CPS use are presented.     

High heat load properties of actively cooled W/CuCrZr mock-ups by diffusion bonding with Ni or Ti interlayer

Two actively cooled mock-ups with 5 mm thick tungsten armor, joined to CuCrZr alloy, were successfully developed by diffusion bonding technique with Ti or Ni interlayer for the EAST device in ASIPP. Its thermal response and thermal fatigue properties were investigated with active cooling. No cracks and voids occurred at the interface of W/CuCrZr after thermal response test with a heat flux from 0 MW/m² to 10 MW/m². It survived up to 200 cycles under 10 MW/m². The temperature distributions of the mock-up were estimated by Finite Element Analysis. The simulation results indicated that thermal contact capability between the tungsten and the copper alloy with Ti interlayer was higher than that of Ni interlayer. Results showed that diffusion bonding of W/CuCrZr with Ni or Ti interlayer is a potential candidate for a high heat resistance armor material on plasma facing components (PFC).     

Non-destructive examination of the bonding interface in DEMO divertor fingers

Plasma facing components (PFCs) with tungsten (W) armor materials for DEMO divertor require a high heat flux removal capability (at least 10 MW/m² in steady-state conditions). The reference divertor PFC concept is a finger with a tungsten tile as a protection and sacrificial layer brazed to a thimble made of tungsten alloy W – 1% La₂O₃ (WL10). Defects may be located at the W thimble to W tile interface. As the number of fingers is considerable (>250,000), it is then a major issue to develop a reliable control procedure in order to control with a non-destructive examination the fabrication processes. The feasibility for detecting defect with infrared thermography SATIR test bed is presented. SATIR is based on the heat transient method and is used as an inspection tool in order to assess component heat transfer capability. SATIR tests were performed on fingers integrating or not the complex He cooling system (steel cartridge with jet holes). Millimeter size artificial defects were manufactured and their detectability was evaluated. Results of this study demonstrate that the SATIR method can be considered as a relevant non-destructive technique examination for the defect detection of DEMO divertor fingers.     

Thermo-mechanical analysis of Wendelstein 7-X plasma facing components

The stellarator experiment Wendelstein 7-X (W7-X) is designed for stationary plasma operation (30 min). Plasma facing components (PFCs) such as the divertor targets, baffles, heat shields and wall panels are being installed in the plasma vessel (PV) in order to protect it and other in-vessel components. The different PFCs will be exposed to different magnitude of heat loads in the range of 100 kW/m²–10 MW/m² during plasma operation. An important issue concerning the design of these PFCs is the thermo-mechanical analysis to verify their suitability for the specified operation phases. A series of finite element (FE) simulations has been performed to achieve this goal. Previous studies focused on the test divertor unit (TDU) and high heat flux (HHF) target elements. The paper presents detailed FE thermo-mechanical analyses of a prototype HHF target module, baffles, heat shields and wall panels, as well as benchmarking against tests.     

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.