This paper presents the results of high heat flux tests at Sandia National Laboratories of mock-ups armored with W rods embedded in water-cooled copper-alloy heat sinks. The major result is the excellent performance of these mock-ups in thermal response tests at up to 30 MW/m2 and in thermal cycling tests of 500 cycles (10 s on, 10 s off) at ∼25 MW/m2. Melting of rod tips and the subsequent ‘self-repair’ of a degraded thermal bond were seen. Issues for further development in both testing and manufacturing of the mock-ups were identified. Evaluation of the surface heat flux was found to be somewhat problematic and this issue is discussed.

Over the past five years, four international parties, which include the European Communities, Japan, the Russian Federation and the United States, have been collaborating on the design and development of the International Thermonuclear Experimental Reactor (ITER), the next generation magnetic fusion energy device. During the ITER Engineering Design Activity (EDA), beryllium plasma spray technology was investigated by Los Alamos National Laboratory as a method for fabricating and repairing and the beryllium first wall surface of the ITER tokamak. Significant progress has been made in developing beryllium plasma spraying technology for this application. Information will be presented on the research performed to improve the thermal properties of plasma sprayed beryllium coatings and a method that was developed for cleaning and preparing the surface of beryllium prior to depositing plasma sprayed beryllium coatings. Results of high heat flux testing of the beryllium coatings using electron beam simulated ITER conditions will also be presented.

Mockups for tests were fabricated and thermal response and thermal fatigue lifetime tests using electron beam facilities were carried out to examine material damage and thermal response of carbon carbon fiber composite (CFC) brazed the oxygen free high conductivity (OFHC) for a local island divertor (LID) plate. Model calculation for thermal response was also carried out to explain the phenomena of experimental results. Thermal response tests of an MFC-1 mockup were performed at the condition that the water flow velocity, pressure and temperature were 1.6–10 m/s, 1.0–4.0 MPa and 20°C, 150°C respectively. The MFC-1 mockup showed good heat removal performance. In the case of a CX-2002U mockup with 10 mm armor thickness, surface temperature is near 1000°C at 10 MW/m2. Therefore, use of CX-2002U as armor material is allowable concerning the maximum surface temperature due to heat flux. A thermal fatigue test of the MFC-1 mockup was also performed. Temperatures increase due to degradation was not observed up to 1000 cycles. Calculation for thermal response was performed using a finite element analysis code. The comparison measured with calculated data presented information of critical heat flux and detachment of the tile.

This paper is a review of the current joining technologies for plasma facing components in the US for the International Thermonuclear Experimental Reactor (ITER) project. Many facilities are involved in this project. All of those facilities are not represented in the authors list but all contributions will be noted throughout the report and in the acknowledgements. Many unique and innovative joining techniques are being considered in the quest to join two candidate armor plate materials (beryllium and tungsten) to a copper base alloy heat sink (Glidcop, Elbrador). These techniques include brazing and diffusion bonding, compliant layers at the bond interface, and the use of diffusion barrier coatings and diffusion enhancing coatings at the bond interfaces. The development and status of these joining techniques will be detailed in this report.

Thermo-mechanical modeling has indicated the desirability of using 'brush' structures (clusters of small filaments or rods) for armor in plasma facing components as a means of reducing stresses at the joint with the heat sink. Fabrication challenges have prevented use of this armor configuration to date. Development work in the US has resulted in methods for fabricating tungsten (W) brush structures suitable for scale-up on the ITER targets and dome. Methods developed use 1.6 mm and 3.2 mm diameter W welding electrode as stock for the armor, and welded metallic honeycomb for fixturing. Three approaches to fabricating said structures are: (1) Cu or Cu/W functionally gradient materials (FGMs) are plasma sprayed to the W brushes followed by diffusion bonding to the Cu alloy heat sink, (2) Cu is cast to the tips of the W brushes followed by diffusion bonding to the Cu alloy heat sink, and (3) W rods are coated with a bonding aid and are directly pressed into and diffusion bonded to the Cu alloy heat sink. All three methods utilize low temperature Cu/Cu diffusion bonding techniques also developed in the US under ITER. The honeycomb core may be left in place to provide indications of armor wear.

During the ITER EDA (Engineering Design Activity), the US Home Team developed improved methods for fabricating tungsten armored plasma facing components. Thermo-mechanical modeling indicated the desirability of using "brush-like" structures (clusters of small filaments or rods) as a means of reducing thermal stresses. The commercial availability of tungsten welding electrode rods (1.6 mm and 3.2 mm diameter) significantly reduced the raw material costs. Three approaches were developed: (1) Cu is plasma sprayed to the W brushes followed by diffusion bonding or e-beam welding the copper backing to the CuCrZr heat sink, (2) Molten Cu is cast directly on the tips of the W brushes followed by the diffusion bonding step, and(3) W rods are sputter-coated with a bonding aid and are directly bonded to the CuCrZr heat sink using HIP or vacuum hot pressing. High heat flux testing was performed up to 18 MW/m(2) without damage to two small-scale divertor mockups.

Plasma-sprayed beryllium ITER first wall mock-ups have survived 3000 thermal fatigue cycles at 1 MW/m2 without damage during testing at the Plasma Materials Test Facility at Sandia National Laboratory in New Mexico. This heat flux level is twice the expected design heat flux for ITER first wall modules. Plasma sprayed beryllium mock-ups were vacuum plasma sprayed at the Los Alamos National Laboratory's Beryllium Atomization and Thermal Spray Facility. Results will be reported on the fabrication, high heat flux testing and post-mortem analysis of two beryllium plasma sprayed mock-ups (1) beryllium plasma sprayed directly on a CuNiBe heat sink and (2) beryllium plasma sprayed on a compliant layer of aluminum which was explosion-bonded to a CuCrZr copper heat sink. The high heat flux tests utilized the 30 kW Electron Beam Test System at a Sandia National Laboratory.

Task T-222 of the International Thermonuclear Experimental Reactor (ITER) program addresses the manufacturing and testing of permanent components for use in the ITER divertor. Thermal-hydraulic and critical heat flux performance of the heat sinks proposed for use in the divertor vertical target are part of subtask T-222.4. As part of this effort, two single channel, medium-scale, bare copper alloy, hypervapotron mock-ups were designed by Sandia National Laboratories and McDonnell Douglas Aerospace (MDA), fabricated at MDA and tested at Sandia' Plasma Materials Test Facility using the EB-1200 electron beam system. The objectives of our effort were to develop the design and manufacturing procedures required for construction of robust HHF components, verify thermal-hydraulic, thermomechanical and CHF performance under ITER relevant conditions, and perform analyses of HHF data to identify design guidelines, failure criteria and possibly modify any applicable CHF correlations. This paper describes the design, fabrication and finite elements modeling of two types of hypervapotrons, a common version already in use at JET and a new attached- fin design. HHF test data on the attached-fin hypervapotron will be used to compare the CHF performance under uniform heating profiles on long heated lengths to that of localized, highly peaked, off-nominal profiles.

A new reciprocating Langmuir probe was used to measure density and temperature profiles, ion flow, and potential fluctuation levels from the lower divertor floor up to the X point on the DIII-D Tokamak. This probe is designed to make fast (2 kHz swept, 20 kHz Mach, 500 kHz Vfloat) measurements with 2 mm spatial resolution in the region where the largest gradients on the plasma open flux tubes are found and therefore provide the best benchmarks for scrap-off layer and divertor numerical models. Profiles are constructed using the 300 ms time history of the probe measurements during the 25 cm reciprocating stroke. Both single and double null plasmas can be measured and compared with a 20 Hz divertor Thomson scattering system. The probe head is constructed of four different kinds of graphite to optimize the electrical and thermal characteristics. Electrically insulated pyrolytic graphite rings act as a heat shield to absorb the plasma heat flux on the probe shaft and are mounted on a carbon/carbon composite core for mechanical strength. The Langmuir probe sampling tips are made of a linear carbon fiber composite. The mechanical, electrical, data acquisition, and power supply systems will be described. Initial measurements will also be presented.

A novel technique has been used to test the relative low cycle thermal fatigue resistance of different grades of US and Russian beryllium, which is proposed as plasma facing armor for fusion reactor first wall, limiter and divertor components. The 30 kW electron beam test system at Sandia National Laboratories was used to sweep the beam spot along one direction at 1 Hz. This produces a localized temperature `spike' of 750°C for each pass of the beam. Large thermal stresses in excess of the yield strength are generated, due to very high spot heat flux, 250 MW m−2. Cyclic plastic strains on the order of 0.6% produced visible cracking on the heated surface in less than 3000 cycles. An in-vacuo fiber optic borescope was used to visually inspect the beryllium surfaces for crack initiation. Grades of US beryllium tested included: S-65C, S-65H, S-200F, S200F-H, SR-200, I-400, extruded high purity, HIP′d spherical powder, porous beryllium (94 and 98% dense), Be/30% BeO, Be/60% BeO, and TiBe12. Russian grades included: TGP-56, TShGT, DShG-200, and TShG-56. Both the number of cycles to crack initiation and the depth of crack propagation, were measured. The most fatigue resistant grades were S-65C, DShG-200, TShGT and TShG-56. Rolled sheet Be (SR-200) showed excellent crack propagation resistance in the plane of rolling, despite early formation of delamination cracks. Only one sample showed no evidence of surface melting, Extruded (T). Metallographic and chemical analyses are provided. Good agreement was found between the measured depth of cracks and a 2-D elastic-plastic finite element stress analysis.

Thermal response and thermal fatigue tests of four 5-mm-thick beryllium tiles on a Russian Federation International Thermonuclear Experimental Reactor (ITER)-relevant divertor mock-up were completed on the electron beam test system at Sandia National Laboratories. The beryllium tiles were diffusion bonded onto an oxygen-free high-conductivity copper saddle-block and a dispersion-strengthened copper alloy tube containing a copper porous coating. Thermal response rests were performed on the files to an absorbed heat flux of 5 MW/m(2) and surface temperatures near 300 degrees C using 1.4 MPa wafer at 5 m/s flow velocity and an inlet temperature of 8 to 15 degrees C. One tile was exposed to incrementally increasing heat fluxes up to 9.5 MW/m(2) and surface temperatures up to 690 degrees C before debonding at 10 MW/m(2). A second tile debonded in 25 to 30 cycles at <0.5 MW/m(2). However, a third file debonded after 9200 thermal fatigue cycles at 5 MW/m(2), while another debonded after 6800 cycles. Posttest surface analysis indicated that fatigue failure occurred in the intermetallic layers between the beryllium and copper. No fatigue cracking of the bulk beryllium was observed. It appears that microcracks growing at the diffusion bond produced the observed gradual temperature increases during thermal cycling. These experiments indicate that diffusion-bonded beryllium files can survive several thousand thermal cycles under ITER-relevant conditions. However, the reliability of the diffusion-bonded joint remains a serious issue.

A novel technique has been used to test the relative low cycle thermal fatigue resistance of different grades of US and Russian beryllium, which is proposed as plasma facing armor for fusion reactor first wall, limiter, and divertor components. The 30 kW electron beam test system at Sandia National Laboratories was used to sweep the beam spot along one direction at 1 Hz. This produces a localized temperature ``spike`` of 750{degree}C for each pass of the beam. Large thermal stresses in excess of the yield strength are generated due to very high spot heat flux, 250 MW/m{sup 2}. Cyclic plastic strains on the order of 0.6% produced visible cracking on the heated surface in less than 3000 cycles. An in-vacuo fiber optic borescope was used to visually inspect the beryllium surfaces for crack initiation. Grades of US beryllium tested included: S-65C, S- 65H, S-200F, S-200F-H, SR-200, I-400, extruded high purity, HIP`d spherical powder, porous beryllium (94% and 98% dense), Be/30% BeO, Be/60% BeO, and TiBe{sub 12}. Russian grades included: TGP-56, TShGT, DShG-200, and TShG-56. Both the number of cycles to crack initiation, and the depth of crack propagation, were measured. The most fatigue resistant grades were S-65C, DShG-200, TShGT, and TShG-56. Rolled sheet Be (SR-200) showed excellent crack propagation resistance in the plane of rolling, despite early formation of delamination cracks. Only one sample showed no evidence of surface melting, Extruded (T). Metallographic and chemical analyses are provided. Good agreement was found between the measured depth of cracks and a 2-D elastic-plastic finite element stress analysis.

Beryllium, carbon-based materials and tungsten are considered as plasma facing materials for the next generation of fusion machines such as the international thermonuclear experimental reactor (ITER). Beryllium is one of the primary candidate materials because of its low atomic number and lack of tritium codeposition. However, joining of a beryllium armor to a copper heat sink remains a critical problem due to the formation of brittle intermetallics at the interface. To address this concern, the Japan Atomic Energy Research Institute manufactured a beryllium/Cu divertor module with Cr and Ni diffusion barriers. This Be/Cu module was tested in the electron beam test system of Sandia National Laboratories in the framework of the US—Japan Fusion Collaboration. The divertor module consisted of four beryllium tiles, 25 mm × 25 mm, and a square copper heat sink with convolutions like a screw nut inside the coolant channel. To evaluate the integrity of the brazed bonds under various heat fluxes, beryllium tiles of two different thicknesses, 2 and 10 mm, were bonded to the copper heat sink. Cooling conditions of 10 m/s water flow velocity at 1 MPa, and a water inlet temperature of 20°C were selected based on the thermal analysis. During high heat flux testing the 10 mm thick Be tiles detached at an absorbed heat flux around 5 MW/m2 for several shots due to flaws at the braze joint confirmed by optical observation after manufacturing. One of the 2 mm thick Be tiles failed after 550 cycles at the steady state heat flux of 6.5 MW/m2. Most likely the failure was caused by brittleness at the interface caused by the presence of BeCu intermetallics.

Plasma-spray technology is under investigation as a method for producing high thermal conductivity beryllium coatings for use in magnetic fusion applications. Recent investigations have focused on optimizing the plasmaspray process for depositing beryllium coatings on damaged beryllium surfaces. Of particular interest has been optimizing the processing parameters to maximize the through-thickness thermal conductivity of the beryllium coatings. Experimental results will be reported on the use of secondary H2 gas additions to improve the melting of the beryllium powder and negative transferred-arc cleaning to improve the bonding between the beryllium coatings and the underlying surface. Information will also be presented on thermal cycle tests which were done on beryllium coated ISX-B beryllium limiter tiles using 10s cycle times with 60s cooldowns using a heat flux slightly in excess of 5 MW/m2.

High heat flux testing for the US fusion power program is the primary mission of the Plasma Materials Test Facility (PMTF) located at Sandia National Laboratory. This facility, an official Department of Energy User Facility, has been in operation for over 15 years and has provided much of the high heat flux data used in the design and evaluation of plasma facing components for many of the world's magnetic fusion tokamak experiments. In addition to domestic tokamaks such as Tokamak Fusion Test Reactor at Princeton, the DIII-D tokamak at General Atomics, and Alcator C-Mod at MIT, components for international experiments like TEXTOR, Tore- Supra, and Jet also have been tested at the PMTF. High heat flux testing spans a wide spectrum including thermal shock tests on passively cooled materials, thermal response and thermal fatigue tests on actively cooled components, critical heat flux burnout testes, braze reliability tests, and safety related tests. The program's main focus now is on testing of beryllium and tungsten armor tiles bonded to divertor, limiter, and first wall components for the International Thermonuclear Experimental Reactor (ITER). The ITER project is a collaboration among the US, EU, RF, and Japanese fusion programs. This article provides a brief overview of the high heat flux testing capabilities at the PMTF, and describes some recent test results.

Several types of porous media heat exchangers are being evaluated for use infusion applications. Broadly, these devices can be classified as capillary-pumped (heat pipes) or mechanically-pumped heat exchangers. Monel/water thermosyphon heat pipes with a porous metal wick are being evaluated for use in Faraday shields. A subscale prototype has been fabricated and initial tests at Oak Ridge National Laboratory have shown favorable results. Alkali metal heat pipes have demonstrated absorbed heat flux capability of over 1000 MW/m(2). An advanced gyrotron microwave cavity is being developed that uses water cooling in a mechanically-pumped copper porous metal heat exchanger. Tests on a prototype demonstrated absorbed heat flux capability in excess of 100 MW/m(2). Porous metal heat exchangers with helium, water, or liquid metal coolants are being evaluated for plasma-facing component cooling. Tests on a helium/copper porous metal heat exchanger demonstrated absorbed heat flux capability in excess of 15 MW/m(2). Applications, conceptual designs, fabricated hardware, and test results are summarized.

The Loss of Flow Accident (LOFA) is a serious safety concern for the International Thermonuclear Experimental Reactor (ITER) as it has been suggested that greater than 100 seconds are necessary to safely shutdown the plasma when ITER is operating at full power. In this experiment, the thermal response of a prototypical ITER divertor tube during a simulated LOFA was studied. The divertor tube was fabricated from oxygen-free high-conductivity copper to have a square geometry with a circular coolant channel. The coolant channel inner diameter was 0.77 cm, the heated length was 4.0 cm, and the heated width was 1.6 cm. The mockup did not feature any flow enhancement techniques, i.e., swirl tape, helical coils, or internal fins. One-sided surface heating of the mockup was accomplished through the use of the 30 kW Sandia Electron Beam Test System. Alter reaching steady state temperatures in the mockup, as determined by two Type-K thermocouples installed 0.5 mm beneath the heated surface, the coolant pump was manually tripped off and the coolant flow allowed to naturally coast down. Electron beam heating continued after the pump trip until the divertor tube's heated surface exhibited the high temperature transient normally indicative of rapidly approaching “burnout”. Experimental data showed that time-to-burnout increases proportionally with increasing inlet velocity and decreases proportionally with increasing incident heat flux

Plasma-spray technology is under investigation as a method for producing high thermal conductivity beryllium coatings for use in magnetic fusion applications. Recent investigations have focused on optimizing the plasma-spray process for depositing beryllium coatings on damaged beryllium surfaces. Of particular interest has been optimizing the processing parameters to maximize the through-thickness thermal conductivity of the beryllium coatings. Experimental results will be reported on the use of secondary H2 gas additions to improve the melting of the beryllium powder and negative transferred-arc cleaning to prepare beryllium surfaces prior to depositing beryllium. Information will also be presented on thermal fatigue tests which were performed on beryllium coated ISX-B beryllium limiter tiles using 10 sec cycle times with 60 sec cooldowns and an International Thermonuclear Experimental Reactor (ITER) relevant divertor heat flux slightly in excess of 5 MW/ml

Plasma-spraying of beryllium is under investigation as a potential method for coating the first wall blanket surface of the International Thermonuclear Experimental Reactor (ITER) and as a technique for regenerating damaged beryllium surfaces as a result of normal and off-normal operating events. Methods to optimize the thermal conductivity of vacuum plasma-sprayed (VPS) beryllium are currently under investigation at the Los Alamos National Laboratory's Beryllium Atomization and Thermal Spray Facility (BATSF). Studies are being conducted to evaluate the effect of the starting beryllium feedstock powder, processing parameters and resulting microstructures on the thermal conductivity of VPS beryllium. The characteristic layered microstructure, typical of plasma-sprayed deposits, contains regions of incomplete bonding between individual splat layers which can decrease the thermal conductivity through the thickness of the deposit. Impurities present in the starting beryllium feedstock powder can also influence the thermal conductivity of VPS beryllium by causing impurity striations throughout the thickness of the deposit.

The behavior of thick B4C coatings on different graphites under high power electron beam irradiation, pulsed plasma irradiation and under DIII-D divertor plasma action was investigated. SiCB4C coating on graphites produces in General Atomics and pyrolytic boron nitride were also tested in the plasma gun device. In the following tests at these facilities, the samples were examined using SIMS, EDAX, X-ray crystallography, profilometry and Auger spectroscopy. B4C coatings showed excellent durability under high heat flux irradiation and in conditions of real tokamak divertor plasma. The obtained results indicated that the use of RGT graphite (the graphite with high thermal conductivity) improves the durability of the B4C coating significantly. Pyrolytic boron nitride showed very small removal of the matter and no mechanical damage.