Design Of Plasma Facing Components Research Articles

Fracture mechanical analysis of a tungsten monoblock-type plasma-facing component without macroscopic interlayer for high-heat-flux divertor target

In the framework of the European DEMO divertor project, several novel design concepts of the plasma-facing components of vertical targets are being developed. One of those concepts is the tungsten monoblock design (similar to the ITER divertor) but with a very thin interlayer (roughly 25mm thick only) at the armor/tube bond interface instead of a thick (1mm) copper interlayer as has been the case in the conventional tungsten monoblock design developed for ITER divertor. The thin interlayer serves as bonding agent, but not as structural constituent. The reasoning for this novel design concept to omit the thick soft copper interlayer, which has been used as stress-relaxing buffer between the stiff armor block and tube, is to prevent plastic fatigue damage under cyclic high heat flux loads and irradiation embrittlement which the soft copper interlayer is predicted to undergo. On the other hand, the desirable stress relaxation effect on a global scale is abandoned. In this study, such trade-off effects are computationally investigated in a comparative assessment of structural impact, which the presence (or absence) of the thick copper interlayer is expected to bring forth, in terms of the fracture and fatigue behaviour of the armour block and cooling tube in two representative cases of tungsten monoblock plasma facing component design, namely, with and without a thick copper interlayer. Quantitative results of cyclic plastic strain history and crack tip fracture energy are presented for the armour surface, bond interface and tube of the respective plasma facing component models. The positive and negative implications of these impacts on the structural integrity are discussed.

Modelling deuterium release from tungsten after high flux high temperature deuterium plasma exposure

Tungsten is a primary candidate for plasma facing materials for future fusion devices. An important safety concern in the design of plasma facing components is the retention of hydrogen isotopes. Available experimental data is vast and scattered, and a consistent physical model of retention of hydrogen isotopes in tungsten is still missing. In this work we propose a model of non-equilibrium hydrogen isotopes trapping under fusion relevant plasma exposure conditions. The model is coupled to a diffusion-trapping simulation tool and is used to interpret recent experiments involving high plasma flux exposures. From the computational analysis performed, it is concluded that high flux high temperature exposures (T = 1000 K, flux = 1024 D/m2/s and fluence of 1026 D/m2) result in generation of sub-surface damage and bulk diffusion, so that the retention is driven by both sub-surface plasma-induced defects (bubbles) and trapping at natural defects. On the basis of the non-equilibrium trapping model we have estimated the amount of H stored in the sub-surface region to be ∼10−5 at−1, while the bulk retention is about 4 × 10−7 at−1, calculated by assuming the sub-surface layer thickness of about 10 μm and adjusting the trap concentration to comply with the experimental results for the integral retention.

Helium segregation on surfaces of plasma-exposed tungsten

We report a hierarchical multi-scale modeling study of implanted helium segregation on surfaces of tungsten, considered as a plasma facing component in nuclear fusion reactors. We employ a hierarchy of atomic-scale simulations based on a reliable interatomic interaction potential, including molecular-statics simulations to understand the origin of helium surface segregation, targeted molecular-dynamics (MD) simulations of near-surface cluster reactions, and large-scale MD simulations of implanted helium evolution in plasma-exposed tungsten. We find that small, mobile Hen (1 ⩽ n ⩽ 7) clusters in the near-surface region are attracted to the surface due to an elastic interaction force that provides the thermodynamic driving force for surface segregation. This elastic interaction force induces drift fluxes of these mobile Hen clusters, which increase substantially as the migrating clusters approach the surface, facilitating helium segregation on the surface. Moreover, the clusters’ drift toward the surface enables cluster reactions, most importantly trap mutation, in the near-surface region at rates much higher than in the bulk material. These near-surface cluster dynamics have significant effects on the surface morphology, near-surface defect structures, and the amount of helium retained in the material upon plasma exposure. We integrate the findings of such atomic-scale simulations into a properly parameterized and validated spatially dependent, continuum-scale reaction-diffusion cluster dynamics model, capable of predicting implanted helium evolution, surface segregation, and its near-surface effects in tungsten. This cluster-dynamics model sets the stage for development of fully atomistically informed coarse-grained models for computationally efficient simulation predictions of helium surface segregation, as well as helium retention and surface morphological evolution, toward optimal design of plasma facing components.

Plasma facing components: a conceptual design strategy for the first wall in FAST tokamak

Satellite tokamaks are conceived with the main purpose of developing new or alternative ITER- and DEMO-relevant technologies, able to contribute in resolving the pending issues about plasma operation. In particular, a high criticality needs to be associated to the design of plasma facing components, i.e. first wall (FW) and divertor, due to physical, topological and thermo-structural reasons. In such a context, the design of the FW in FAST fusion plant, whose operational range is close to ITER’s one, takes place. According to the mission of experimental satellites, the FW design strategy, which is presented in this paper relies on a series of innovative design choices and proposals with a particular attention to the typical key points of plasma facing components design. Such an approach, taking into account a series of involved physical constraints and functional requirements to be fulfilled, marks a clear borderline with the FW solution adopted in ITER, in terms of basic ideas, manufacturing aspects, remote maintenance procedure, manifolds management, cooling cycle and support system configuration.

Thermal-hydraulic and thermo-mechanical design of plasma facing components for SST-1 tokamak

The Plasma Facing Components (PFCs) are one of the major sub-systems of SST-1 tokamak. PFC of SST-1 consisting of divertors, passive stabilizers, baffles and limiters are designed to be compatible for steady state operation. The main consideration in the design of the PFC cooling is the steady state heat removal of up to 1 MW/m2. The PFC has been designed to withstand the peak heat fluxes and also without significant erosion such that frequent replacement of the armor is not necessary. Design considerations included 2-D steady state and transient tile temperature distribution and resulting thermal loads in PFC during baking, and cooling, coolant parameters necessary to maintain optimum thermal-hydraulic design, and tile fitting mechanism. Finite Element (FE) models using ANSYS have been developed to carry out the heat transfer and stress analyses of the PFC to understand its thermal and mechanical behaviors. The results of the calculation led to a good understanding of the coolant flow behavior and the temperature distribution in the tube wall and the different parts of the PFC. Thermal analysis of the PFC is carried out with the purpose of evaluating the thermal mechanical behavior of PFCs. The detailed thermal-hydraulic and thermo-mechanical designs of PFCs of SST-1 are discussed in this paper.

High heat flux performance of brazed tungsten macro-brush test mock-up for divertors

Plasma facing components (PFCs) of divertor will be exposed to steady state and transient heat loads up to 20MW/m2, during operation of ITER-like plasma fusion device. The critical task in fusion research is to design, fabricate and test of PFCs. To withstand high heat loads, PFCs are designed and fabricated in flat tile, mono-block type geometries using tungsten as plasma facing material and CuCrZr alloy is used as a heat sink. These fabricated mock-ups are tested under thermal cyclic heat loads using intense electron beam in pulsed mode. Tungsten macro-brush type of mock-up has been developed by vacuum furnace brazing route. Mock-up was tested to the absorbed heat flux in the range of 0.5–9MW/m2. Simulation of high heat flux (HHF) test under steady state and cyclic heat loads has been done using ANSYS12 finite element analysis (FEA) software. HHF tests have been successfully performed on the tungsten mock-up.

Thermomechanical Analysis of Plasma-Facing Components for SST-1 Tokamak

Plasma-facing components (PFCs) are an important part of the Indian Steady State Superconducting Tokamak (SST-1) design. The main consideration in the design of PFCs is steady-state heat removal of up to 1 MW/m2, which is nearly the limit for incident heat flux of mechanically attached graphite tiles for tokamak PFCs. SST-1 PFCs consist of divertors, passive stabilizers, baffles, and limiters and are designed for long-pulse operation, which requires active cooling of these components. During steady-state operation, the average heat loads on the divertor and passive stabilizers are expected to be 0.6 and 0.25 MW/m2, respectively. Design considerations include two-dimensional steady-state and transient tile temperature distribution and the resulting thermal loads in PFCs during plasma operation. Thermal analysis is carried out to evaluate the thermomechanical behavior of the SST-1 PFCs. In this technical note, temperature distribution and thermally induced stresses and strains in PFCs are analyzed using a finite element method, and the effect of stress and strain on different materials used in PFCs is discussed.

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.

Towards the development of workable acceptance criteria for the divertor CFC monoblock armour

The plasma-facing components (PFCs) of the ITER divertor will be subjected to high heat flux (HHF). Carbon–fibre composite (CFC) is selected as the armour for the region of highest heat flux where the scrape-off layer of the plasma intercepts the vertical targets (VT). Failure of the armour to heat sink joints will compromise the performance of the divertor and could ultimately result in its failure and the shut down of the ITER machine. There are tens of thousands of CFCs to CuCrZr joints. The aim of the PFC design is to ensure that the divertor can continue to function even with the failure of a few joints. In preparation for writing the procurement specification for the ITER vertical target PFCs, a programme of work is underway with the objective of defining workable acceptance criteria for the PFC armour joints.

Thermal analysis of plasma facing components of SST-1 Tokamak by finite difference and finite element methods

Thermal analysis is a prime consideration in the design of plasma facing components (PFC) cooling during plasma operation in a Tokamak device. The task is greatly simplified by using computer-based numerical techniques. Finite difference method (FDM) and finite element method (FEM) are the two methods generally being used for thermal analysis. FEM is gaining wide acceptance for such problems because it performs all the necessary computations in the computer. FDM, in contrast, requires significant amounts of data to be calculated for input into the computer program. This can provide accurate thermal analysis in less time and less computer memory than FEM. An efficient finite difference (FD) code has been developed to derive the two-dimensional temperature profile in different PFC modules subjected to steady-state condition. The details of the code and comparison of its results with finite element (FE) analysis are presented in this paper.

Design of plasma facing components for the ITER feat divertor

A comprehensive design of the ITER divertor has been developed within the EU R&D for ITER. It consists of plasma facing components (PFCs) and cassettes body (CB). The PFCs are actively cooled thermal shields while the CB are massive supports for the PFCs providing also a neutronic shield. The present paper gives a detailed design of the PFCs and the CBs. It includes the cooling path, the manifolds, the attachments between the PFCs and the CBs and those between the CBs and the vacuum vessel (VV). The design has been carried out with a series of analyses. The neutronics analysis assures the shielding efficiency of the PFCs/CBs towards the rear components, and calculates the radiation loads in the key points. The electromagnetic (EM) analysis evaluates the loads due to eddy and halo currents. The thermo–hydraulic analysis verifies the effectiveness of the cooling circuit regarding the minimum margin critical heat flux (CHF) and the maximum acceptable pressure drop. The thermo-mechanical analysis verifies the integrity of the components under coolant pressure, neutron/plasma thermal loads and EM loads. The work, done with the collaboration of professionals from different organisations, shows that the proposed design of the components fulfils all the requirements of the ITER FEAT machine.

Engineering design and thermal hydraulics of plasma facing components of SST-1

SST-1 is a medium size tokamak with super conducting magnetic field coils. All the subsystems of SST-1 are designed for quasi steady state (≈1000 s) operation. Plasma Facing Components (PFCs) of SST-1 consisting of divertors, passive stabilizers, baffles and poloidal limiters are also designed to be compatible for steady state operation. As SST-1 is designed to run double null divertor plasmas, these components also have up-down symmetry. A closed divertor configuration is chosen to produce high recycling and high pumping speed in the divertor region. All the PFC are made of copper alloys (CuCrZr and CuZr) on which graphite tiles are mechanically attached. These copper alloy back plates are actively cooled with water flowing in the channels grooved on them with the main consideration in the design of PFCs as the steady state heat removal of about 1.0 MW/m 2. In addition to be able to remove high heat fluxes, the PFCs are also designed to be compatible for baking at 350 °C. Extensive studies, involving different flow parameters and various cooling layouts, have been done to select the final cooling parameters and layout. Thermal response of the PFCs and vacuum vessel during baking, has been calculated using a fortran code and a 2-D finite element analysis. The PFCs and their supports are also designed to withstand large electro-magnetic forces. Finite element analysis using ansys software package is used in this and other PFCs design. The engineering design including thermal hydraulics for cooling and baking of all the PFCs is completed. Poloidal limiters are being fabricated. The remaining PFCs, viz. divertors, stabilizers and baffles are likely to go for fabrication in the next few months. The detailed engineering design, the finite element calculations in the structural and thermal designs are presented in this paper.

K-2106 核融合動力炉A-SSTR2プラズマ対向機器の熱構造設計(S27-2 次世代エネルギー源に対応した発電技術(2))(S27 次世代エネルギー源に対応した発電技術)

K-2106 核融合動力炉A-SSTR2プラズマ対向機器の熱構造設計(S27-2 次世代エネルギー源に対応した発電技術(2))(S27 次世代エネルギー源に対応した発電技術)

Design issues and cost implications of RTO/RC-ITER divertor

This paper reports on the conceptual divertor design developed for the reduced technical objectives/reduced cost-international thermonuclear experimental reactor (RTO/RC-ITER). The cost drivers are discussed and a number of cost-reducing measures identified. Scaled costs, based on industrial estimates of the 1998 ITER design (Technical Basis for the ITER Final Design Report, Cost Review and Safety Analysis (FDR). ITER EDA Documentation Series No. 16. IAEA, Vienna, 1998), give for the RTO/RC-ITER ≈60% of the FDR costs. Plasma facing components (PFCs) account for 75% of the total divertor costs. Hence, PFC design simplifications are outlined in the paper showing the possibility of achieving a cost reduction of 50%. The design proposals, outlined in the paper, focus on minimising the number of sub-components and simplifying the manufacturing cycle. These changes contribute to improved reliability based on a more robust coolant design layout. The reduced space allocated to the divertor (G. Janeschitz, A. Antipenkov, V. Barabash, S. Chiocchio, G. Federici, C. Ibbott, E. Martin, R. Tivey, Overview of the Divertor Design and its Integration into RTO/RC-ITER, this conference) requires changes to the design that minimise the cassette body thickness, relocate the cassette attachments and revise the remote handling philosophy. Results of supporting electro-magnetic, neutron shielding, thermo-hydraulic and pumping conductance analyses are reported, qualifying the cassette design. A reduction in the coolant inlet temperature to 100–120°C is discussed in terms of thermal-hydraulic performance and fatigue life of the heat sink. Finally, an R&D plan sets out the work needed: (1) to develop the cost saving measures of the new design; and (2) to demonstrate the reliability of the chosen technologies.

Modelling of particle flows in a scrape-off layer with limiter biasing

The superconducting tokamak Tore Supra will be equipped with an actively cooled toroidal pump limiter (TPL), in the framework of the CIEL (Composants Internes Et Limiteurs) project, dedicated to plasma facing component design for steady state operation. The TPL is equipped with throats, located only on the high field side, for particle collection allowing the control of plasma density which is essential for long plasma discharges. The present design work of the CIEL includes a biasing system in order to enhance the particle pumping. A fluid model, based on the classical fluid equation, is used to estimate the effects of the electric field on the particle flows in the Scrape-Off Layer (SOL). The modifications of the density, the particle flow (toroidal and poloidal) and the position of the stagnation point are discussed as a function of the bias voltage. The model clearly illustrates the different resulting effects on particle pumping for a divertor and a limiter configuration which are designed respectively for poloidal or parallel particle collection.

The thermal shock resistance of a joining material of C/C composite and copper

Plasma facing materials for the next fusion reactor devices will have severe problems such as thermal shock fracture, surface erosion and injection of sputtering particles to plasma caused by charged particle fluxes and very high heat shocks. A joining material, in which a carbon fiber reinforced carbon composite was joined to oxygen-free copper by inserting a molybdenum plate and some metallic films in the joining layers, was developed for a solution of those problems. The thermal shock resistance and the thermal shock fracture toughness were evaluated by an eccentric local heating method of arc discharge. The joining material did not fracture during severe thermal shock tests such as plasma disruption, however, thermal and delamination cracks were observed at the joining parts by scanning electron microscope (SEM). These results can be useful in contributing to the development and the safety design of plasma facing components for fusion reactor devices.

Design of the ITER EDA plasma facing components

The design of the plasma facing components (PFC) in ITER has evolved with the detailed design of the reactor. The structures exposed to the plasma have different requirements according to their functions. The primary wall, surrounding most of the plasma along the last closed magnetic surface, is exposed to a moderate heat flux (0.5 MW m −2) but has to withstand the highest neutron load. The baffle wall is exposed to a peak heat flux of 3 MW m −2 and to severe erosion from neutral particles due to their high neutrals pressure in the divertor. The limiter is subjected to the same loads as the primary wall during plasma burn conditions and a higher peak heat flux (depending on its location) during the start-up and shut down phases when the plasma is leaning on its surface. The divertor vertical targets intercept the open magnetic flux surfaces near the separatrix and have to withstand the highest heat flux and erosion in their lower part. The divertor dome is located directly below the null point and works in conditions similar to the baffle. The divertor wings receive similar thermal loads as the dome but can be subjected to high heat shocks and electromagnetic forces during plasma disruption. The paper describes the solutions adopted for the PFC and the results of analyses performed to validate the design. The description is focused on the part of the PFC which is exposed to the plasma.

Thermal Shock Resistances of a Joining Material of C/C Composite and Copper.

The purpose of this study is to contribute to the development and safety design of plasma facing components for fusion reactor devices. We evaluated the thermal shock resistance and the thermal shock fracture toughness of a joining material which was used to join a carbon fiber reinforced carbon composite(C/C composite) to oxygen-free copper. We also examined the microstructures of the joining layers using a scanning electron microscope before and after thermal shock tests. The joining material did not fracture during the thermal shock tests. However, thermal cracks and delamination cracks were observed in the joining layers.

Possible divertor solutions for a fusion reactor.: Part I. Physical aspects based on present day divertor operation

With an anticipated power flux across the separatrix of up to 300 MW of an ITER-like fusion reactor, conventional measures of power spread lead to a peak power load at the target plates in the order of 30 MW m −2, far beyond the technically feasible limit for stationary operation. Radiative cooling by seed impurities appears to be the most promising plasma-physical option to reduce the target power load, but extrapolations of present experiments predict an only marginally tolerable increase of the plasma effective charge Z eff. Key points will be the achievement of very high electron densities, leading to more effective radiative cooling by δ P rad/ δ Z eff∝ n e 2 while keeping the edge temperature within its optimum range. This range is bounded from below by the H→L mode temperature threshold due to confinement requirements, whereas the upper boundary is given by the ideal ballooning stability limit which is connected to type-I ELM activity which may cause non-tolerable divertor heat loads. The completely detached H-mode (CDH) in ASDEX Upgrade demonstrates radiative H-mode operation within this operational range exhibiting high-frequent type-III ELMs and target power load in the order of 10% of the heating power. At present, open questions on high density reactor operation are related to radiative instabilities as well as edge transport enhancement and H-mode impairment observed in several tokamaks under high density conditions. Measures to overcome these detrimental effects will be investigated with improved divertor concepts in the near future. The possible problems connected to high density reactor operation can be relaxed, if the design of plasma facing components with higher heat flux endurance is successful.

RACLETTE: a model for evaluating the thermal response of plasma facing components to slow high power plasma transients. Part II: Analysis of ITER plasma facing components

The transient thermal model RACLETTE (acronym of Rate Analysis Code for pLasma Energy Transfer Transient Evaluation) described in part I of this paper is applied here to analyse the heat transfer and erosion effects of various slow (100 ms-10 s) high power energy transients on the actively cooled plasma facing components (PFCs) of the International Thermonuclear Experimental Reactor (ITER). These have a strong bearing on the PFC design and need careful analysis. The relevant parameters affecting the heat transfer during the plasma excursions are established. The temperature variation with time and space is evaluated together with the extent of vaporisation and melting (the latter only for metals) for the different candidate armour materials considered for the design (i.e., Be for the primary first wall, Be and CFCs for the limiter, Be, W, and CFCs for the divertor plates) and including for certain cases low-density vapour shielding effects. The critical heat flux, the change of the coolant parameters and the possible severe degradation of the coolant heat removal capability that could result under certain conditions during these transients, for example for the limiter, are also evaluated. Based on the results, the design implications on the heat removal performance and erosion damage of the variuos ITER PFCs are critically discussed and some recommendations are made for the selection of the most adequate protection materials and optimum armour thickness.