Armor Tiles Research Articles

The effects of radiation damage on the properties of GraphNOL N3M

GraphNOL N3M (N3M) is a bulk graphite developed at Oak Ridge National Laboratory (ORNL) for advanced structural applications in aerospace thermal protection systems. It is currently of interest to the United States fusion energy community for plasma facing components, such as the first wall armor tiles of the International Thermonuclear Experimental Reactor (ITER) because of its superior thermal shock resistance and irradiation lifetime. This paper reports the results of irradiation experiments on N3M graphite at two temperatures, 600 and 875 °C in the High Flux Isotope Reactor (HFIR) at ORNL. Maximum fluences of 4.2 × 10 26 and 2 × 10 26 n/m 2 ( E > 50 keV) or 28.4 and 13.5 dpa (graphite) were attained at 600 and 875° C, respectively. Data are presented on the dimensional stability, volume change, strength, Young's modulus, critical stress intensity factor K Ic, and coefficient of thermal expansion (CTE). The thermal shock resistance of GraphNOL N3M is discussed and comparisons made with other graphites. The influence of irradiation damage on thermal shock resistance is postulated.

Development of high conductive C/C composite tiles for plasma facing armor

C/C composites with high thermal conductivity were developed in unidirectional, two-dimensional and felt types, and were fabricated as full-scale armor tile. Their thermal conductivity in the direction perpendicular to the plasma-side surface is 250 – 550 W/m°C, that is comparable to that of pyrolytic graphite. It was shown by heat load tests that the C/C composites have low surface erosion characteristics and high thermal shock resistance. Various kinds of C/C composites were successfully bonded to metal substrate, and their mechanical strength and thermal shock resistance were tested.

Structures for handling high heat fluxes

The divertor is reconized as one of the main performance limiting components for ITER. This paper reviews the critical issues for structures that are designed to withstand heat fluxes > 5 MW/m 2. High velocity, sub-cooled water with twisted tape inserts for enhanced heat transfer provides a critical heat flux limit of 40–60 MW/m 2. Uncertainties in physics and engineering heat flux peaking factors require that the design heat flux not exceed 10 MW/m 2 to maintain an adequate burnout safety margin. Armor tiles and heat sink materials must have a well matched thermal expansion coefficient to minimize stresses. The divertor lifetime from sputtering erosion is highly uncertain. The number of disruptions specified for ITER must be reduced to achieve a credible design. In-situ plasma spray repair with thick metallic coatings may reduce the problems of erosion. Runaway electrons in ITER have the potential to melt actively cooled components in a single event. A water leak is a serious accident because of steam reactions with hot carbon, beryllium, or tungsten that can mobilize large amounts of tritium and radioactive elements. If the plasma does not shutdown immediately, the divertor can melt in 1–10 s after a loss of coolant accident. Very high reliability of carbon tile braze joints will be required to achieve adequate safety and performance goals. Most of these critical issues will be addressed in the near future by operation of the Tore Supra pump limiters and the JET pumped divertor. An accurate understanding of the power flow out of edge of a DT burning plasma is essential to successful design of high heat flux components.

Engineering design and analyses of the ALT-II particle collection scoops and deflector plates

The magnetic field line tracing and heat transfer analyses required for the design of the particle collection scoops and deflector plates for the Advanced Limiter Test-II (ALT-II) are described, and special fabrication techniques are discussed. For the magnetic field line analysis, expressions for the poloidal and toroidal magnetic fields based on Shafranov stability criteria were used to define areas located at the back of the blade which are shadowed by adjacent deflector plates. The outline of the shadowed areas was then used to contour the sides of the particle collection scoops. Finite element thermal analyses were used to determine the response of the deflector plates, and to adjust the design for the anticipated incident peak heat fluxes of 3000 W/cm 2. Based on the thermal analyses, a 2 mm overhang of the armor tiles over the edge of the blade was selected, and pyrolytic graphite was chosen as the deflector plate material. The computer aided design and fabrication methods used to complete assemblies, including a transfer of computer-aided-design data to numerical control machining tapes, are discussed, and some of the experience gained is summarized.

Infrared thermography system on DIII-D

Six infrared cameras measure temperature changes on the protective graphite armor inside the DIII-D vacuum vessel. Simultaneous time-dependent temperature measurements are made on armor tiles located on the centerpost and divertor regions, and on both outboard limiters. The nearly complete poloidal coverage is useful in measuring both the plasma heat flux distributions inside the vessel and the plasma power balance. Spatial resolution of each camera system is ≲1 cm, while the minimum resolvable time is 125 μs. Data from the IR TV systems are recorded on video tape, and are post-processed serially, using an image processor with an AT-compatible microcomputer. The processing system controls all VCRs, interprets DIII-D timing pulses, digitizes video data in the predetermined regions of interest, averages digitized signals to reduce noise, and constructs data files which are then stored as part of the permanent shot record.

Infrared thermography system on DIII-D (abstract)

Six infrared cameras measure temperature changes on the protective graphite armor inside the DIII-D vacuum vessel. Simultaneous time-dependent temperature measurements are made on armor tiles located on the centerpost and divertor regions, and on both outboard limiters. The nearly complete poloidal coverage is useful in measuring both the plasma heat flux distributions inside the vessel and the plasma power balance. Spatial resolution of each camera system is ≲1 cm, while the minimum resolvable time is 125 μs. Data from the IR TV systems are recorded on video tape, and are post-processed serially, using an image processor with an AT-compatible microcomputer. The processing system controls all VCRs, interprets DIII-D timing pulses, digitizes video data in the predetermined regions of interest, averages digitized signals to reduce noise, and constructs data files which are then stored as part of the permanent shot record.

Behaviour of light impurities in beam-heated jt-60 plasmas with hot graphite walls

The behaviour of light impurities in beam-heated JT-60 plasmas, after changing the limiter and armour tile material from TiC coated molybdenum and inconel to graphite, is discussed. In limiter discharges with high power neutral injection of greater than 12 MW and small gas puff rate, it is generally observed that the influx of carbon increases during the neutral beam phase. This phenomenon is delayed when the base temperature of the vacuum vessel is low. If the heat load to the wall is taken to be 6 MW/m 2 then it is estimated that the increased carbon influx occurs at a wall temperature of 800 K. This temperature coincides with that for maximum chemical sputtering. The strongly rising influx of carbon is therefore attributed to chemical sputtering.

Development, installation, and initial operation of DIII-D graphite armor tiles

An upgrade of the DIII-D vacuum vessel protection system has been completed. The ceiling, floor, and inner wall have been armored to enable operation of CIT-relevant double-null diverted plasmas and to enable the use of the inner wall as a limiting surface. The all-graphite tiles replace the earlier partial coverage armor configuration which consisted of a combination of Inconel tiles and graphite brazed to Inconel tiles. A new all-graphite design concept was chosen for cost and reliability reasons. The ten minute duration between plasma discharge required the tiles to be cooled by conduction to the water-cooled vessel wall. Using two and three-dimensional analyses, the tile design was optimized to minimize thermal stresses with uniform thermal loading on the plasma-facing surface. Minimizing the stress around the tile hold-down feature and eliminating stress concentrators were emphasized in the design. The design of the tile fastener system resulted in sufficient hold-down forces for good thermal conductance to the vessel and for securing the tile against eddy current forces. The tiles are made of graphite, and a program to select a suitable grade of graphite was undertaken. Initially, graphites were compared based on published technical data. Graphite samples were then tested for thermal shock capacity in an electron beam test facility at the Sandia National Laboratory (SNLA) in Albuquerque, New Mexico, USA. The expected operational heat load to the tiles is 500 W/cm 2 for 10 s over the entire tile during limiter operation and 1000 W/cm 2 over a 2 cm wide band for 10 s for divertor operation. Comparative stress analysis was performed on the basis of 1000 W/cm 2 uniformly distributed on the front surface for 5 s with 10 min between pulses. Prototype tiles survived beam heating to the limit of the beam system which was 1000 W/cm 2 for 4 s or 1750 W/cm 2 for 1.75 s. The electron beam facility at the Sandia National Laboratory was used to deposite higher heat flux in a 2 cm wide band in order to model divertor heat flux conditions. These tests were successful with heat fluxes up to 2500 W/cm 2 for 5 s. Both the ion beam and the electron beam tests produced surface temperatures up to 3100 °C which resulted in minor surface erosion but no structural cracking. Installation of approximately 1600 armor tiles was accomplished in November/December 1987. Initial operation experience with the new armor system is summarized.

High heat load tests of a graphite armor first wall with a water cooling jacket

Heat load tests are performed on actively cooled graphite/stainless steel laminated structures, for the simulation of steady-state first wall operations in nuclear fusion reactors. Graphite armor tiles (40 mm × 40 mm × 10 mm) are brazed to stainless steel substrates (1 mm thick, 68 mm diameter) with the insertion of copper-carbon fiber composite (Cu-C) compliant layers (1.7 mm thick). Ag-28 wt%Cu-5wt%Ti solder and Ag-28wt%Cu solder are used for brazing the graphite/ Cu-C interface and Cu-C/stainless steel interface, respectively. Fine grain isotropic graphite and carbon fiber felt reinforced carbon composite (felt C-C) are chosen for armor tile materials. Thermal conductivities of these graphite materials, being measured through a laser flashing method, are 109 and 170 W/mK at room temperature, respectively. Heat load tests are carried out by using a 40 kW DC power source (maximum current: 1000 A). Pulsed thermal loads at heat fluxes ranging from 1 to 6 MW/m2 are realized either through resistance heating of a carbon felt heater on test specimens (1–4 MW/m2) or through electric arcing between a cathode graphite rod and graphite tiles of test specimens (> 4 MW/m2). Surface temperatures of graphite tiles are found to reach equilibrium temperatures at around 20 s after the start of heating. The equilibrium temperatures are higher by 30–40% for isotropic graphite tiles as compared to felt C-C tiles: 1800 °C for isotropic graphite and 1300 °C for felt C-C, at 6 MW/m2. The equilibrium temperature at the Cu-C layer for a 6 MW/m2 heat flux is measured to be around 600 °C.

Divertor and first wall armor design for the TIBER-II engineering test reactor

TIBER-II is a compact, high-power-density, steady-state current-drive engineering test reactor that uses a double-null divertor configuration, operating in the high recycle mode with low particle temperature. The nominal peak heat flux is 3.3 MW/m 2 , with off-normal condition peaks of up to 6.6 MW/m 2 . Plasma disruptions may cause peak energy deposition of up to 13 MJ/m 2 . The design uses water-cooled copper alloy tubes with brazed-on protective tiles. During early operating phases when disruptions may be frequent, these tiles will be made of carbon. During the nuclear testing phase, disruptions must be limited so that more erosion-resistant tungsten tiles may be used. The first walls of TIBER-II experience a heat flux of 0.23 MW/m 2 with disruption energy density of 2.4 MJ/m 2 . They are protected by carbon-carbon composite armor tiles. The TIBER-II design requirements are challenging, and although the divertor and first wall armor designs successfully meet these requirements, significant issues must be resolved to verify their performance. These critical issues and the R&D required to resolve them are described.

Surface damage in TiC coating layers on PDX wall armor tiles

Surface damage and material redepositions are investigated on 17 graphite tile surface of the PDX wall armor exposed to more than 6500 shots of neutral beam injection (NBI) shine-through and more than 700 shots of direct NBI. Titanium deposition layers, from 0.1 to 7 μm in thickness, redeposited metal droplets, or micro-projections with diameters of less than 2–3 μm and a number density at around 10 8/cm 2, are observed on the tile front faces. Aligned wedge-shaped pits are observed at all micro-projections, which are likely due to shadowed Ti-deposition at redeposited metal droplets by Ti-getter evaporation. Blistering observed in both the TiC coating and Ti-deposition layers is due to plasma shine-through D 0 beam irradiations to a fluence of more than 2 × 10 19 at./cm 2.

Properties Required for Isotropic Graphite As First Wall Component of Fusion Device

An isotropic graphite has been widely used for the first wall components of fusion devices as a high heat flux and stable material to plasmas. Since the atomic number of the carbon is low, the radiation loss from the plasma can be largely suppressed compared with metal impurities such as iron.It is known that the primary limiter and the armor tiles of Doublet III and the primary limiter of TFTR were made of POCO graphite and that those components were stable for the plasmas.It is required that the graphite used as the first wall components has low rates for outgassing, recycling and sputtering, high sublimation temperature and higher mechanical strength.This report describes the methods to satisfy these requirements, and examines the heat fluxes to the limiter and the armor tiles.

Modifications of thin coatings by heat treatment: Behavior of redeposited material on the first wall of fusion devices during plasma wall interaction

In magnetic confinement experiments the first wall and especially its high heat flux components (e.g., limiter, armor tiles) are subjected to plasma wall interaction which will cause physical and chemical erosion. The eroded wall material can enter the plasma where it contributes to plasma radiation losses especially when materials with medium or high atomic numbers (Z) are used. To avoid or reduce these objectionable processes, the application of thin low‐Z coatings for the first wall has been discussed and is occasionally already in use. A considerable disadvantage of thin films is their morphological instability under thermal loading. The stability of thin metallic coatings on different substrates was investigated in laboratory experiments under fusion‐relevant heat loading conditions and also under realistic conditions in magnetic confinement experiments.

Post-exposure analysis of C + SiC alloy-coated graphite armor tiles after high heat loadings in a plasma environment

Metallographic and surface microchemical analyses have been performed on C + SiC alloy-coated graphite armor tiles exposed to ∼20 000 plasma discharges and neutral beam pulses of ∼0.2 kW/cm 2 power density with pulse durations up to 300 ms in normal operation in the Doublet III tokamak. In addition, analyses are reported on tiles exposed to 18 neutral beam pulses of ∼2.3 kW/cm 2 with deposition times of ∼150 ms during post-plasma operation to determine tile failure modes. The tiles, acting as limiters, protected the vacuum vessel wall in Doublet III from direct neutral beam heating and performed satisfactorily during the entire operational history. The C + SiC alloy coating, used to prevent erosion of the graphite from chemical sputtering, exhibited no degradation over ∼90% of the 60 tile array. The localized degradation features of the C + SiC alloy coating included arc tracks, micro- and macrocracking, melting, erosion, and foreign metallic deposits and interdiffusion. Extensive metallic deposits in the form of droplets and thin films, and substantial micro- and macrocracking were evident on the alloy coating near limiter ridge regions experiencing the highest heat fluxes. The coating exposed to the high power beam pulses exhibited some erosion.

Maintainability Features of the Compact Ignition Tokamak

The Compact Ignition Tokamak (CIT) is a deuterium-tritium (D-T) device envisaged to be the next experimental reactor in the US Fusion Program. The reactor will initially operate in a nonactivated hydrogen phase for approximately two years. This will permit verification of the integrity of the total system and allow hands-on repair to equipment which has experienced shakedown and early operation failures. Once D-T operations commence, reactor maintenance will require remote handling techniques. An evaluation has been completed to determine what maintenance operations must be performed on the CIT. A maintenance philosophy has been developed which is based upon the use of manipulator systems and robotics in the test cell. Replacement of life-limited equipment will be accomplished using a modular design approach for components, with simple remotely operable interfaces. Examples of operations to be done remotely include: (1) replacing of rf antennae and Faraday shields, (2) uncoupling diagnostic and fueling penetrations, (3) removing of all port covers, and (4) replacing first wall armor tiles, optical mirrors, and vacuum windows.

Residual stresses in bonded armor tiles for in-vessel fusion components

Residual stresses in bonded armor tiles for in-vessel fusion components

Residual stresses in bonded armor tiles for in-vessel fusion components

The residual stresses in a bonded tile/substrate structure were analyzed using both analytical and finite element methods. Beam theory and 2-D elasticity solutions were compared and the latter was found to be more accurate, due to inadequate boundary conditions used in beam theories. Agreement between variational elasticity and finite element solutions was favorable, but the increased flexibility of finite element codes makes them superior when non-linear problems are considered. The response of the calculated stress states to changes in various model dimensions and material parameters was studied parametrically. In general, dimensional changes were found to be significant only for short, thin tiles.

267. The performance of silicon-alloyed pyrolytic carbon coatings on fusion armor and limiter tiles

267. The performance of silicon-alloyed pyrolytic carbon coatings on fusion armor and limiter tiles

Performance results on a C-SiC alloy coating chemically vapor-deposited onto graphite substrates

A new coating material under development by GA Technologies Inc. (GA) was successfully tested in both the limiter and the beam armor tile positions in Doublet III, the experimental tokamak fusion device operated at GA. The coating material is an alloy of about 10 wt.% SiC in pyrolytic carbon deposited onto graphite substrate tiles of sizes up to 10 cm × 10 cm × 2.5 cm. Performance of the coating on graphite tiles under conditions of localized thermal shock was also investigated in laboratory simulation studies using electron beam energy depositions at power densities greater than levels normally experienced in tokamak service. The results of the tests are presented along with observations made during surface examination and microstructural evaluation of the coating. The coating features and deposition model are also described. In a review of related studies of the C-SiC alloy coating, emphasis is placed on the two key advantages, low atomic number and low chemical sputtering, offered by this coating for use on graphite substrates to achieve improved plasma performance in both present and next-generation fusion devices.

Process evaluation and characterization of TiC coating on graphite for doublet III limiters and neutral beam armor

An investigation was conducted to evaluate a thin coating of TiC deposited on a graphite substrate by a commercial CVD coating process. Microstructural analyses are presented for specimens positioned at various locations within the retort during the CVD coating runs. Coating requirements for thickness, quality and uniformity were met. These findings provide the data base for production coating graphite neutral-beam armor and limiter tiles for Tokamaks.