Physics conclusions in support of ITER W divertor monoblock shaping

The key remaining physics design issue for the ITER tungsten (W) divertor is the question of monoblock (MB) front surface shaping in the high heat flux target areas of the actively cooled targets. Engineering tolerance specifications impose a challenging maximum radial step between toroidally adjacent MBs of 0.3 mm. Assuming optical projection of the parallel heat loads, magnetic shadowing of these edges is required if quasi-steady state melting is to be avoided under certain conditions during burning plasma operation and transiently during edge localized mode (ELM) or disruption induced power loading. An experiment on JET in 2013 designed to investigate the consequences of transient W edge melting on ITER, found significant deficits in the edge power loads expected on the basis of simple geometric arguments, throwing doubt on the understanding of edge loading at glancing field line angles. As a result, a coordinated multi-experiment and simulation effort was initiated via the International Tokamak Physics Activity (ITPA) and through ITER contracts, aimed at improving the physics basis supporting a MB shaping decision from the point of view both of edge power loading and melt dynamics. This paper reports on the outcome of this activity, concluding first that the geometrical approximation for leading edge power loading on radially misaligned poloidal leading edges is indeed valid. On this basis, the behaviour of shaped and unshaped monoblock surfaces under stationary and transient loads, with and without melting, is compared in order to examine the consequences of melting, or power overload in context of the benefit, or not, of shaping. The paper concludes that MB top surface shaping is recommended to shadow poloidal gap edges in the high heat flux areas of the ITER divertor targets

First Wall, Blanket, Shield Engineering Technology Program

The First Wall/Blanket/Shield Engineering Technology Program sponsored by the Office of Fusion Energy of DOE has the overall objective of providing engineering data that will define performance parameters for nuclear systems in advanced fusion reactors. The program comprises testing and the development of computational tools in four areas: (1) thermomechanical and thermal-hydraulic performance of first-wall component facsimiles with emphasis on surface heat loads; (2) thermomechanical and thermal-hydraulic performance of blanket and shield component facsimiles with emphasis on bulk heating; (3) electromagnetic effects in first wall, blanket, and shield component facsimiles with emphasis on transient field penetration and eddy-current effects; (4) assembly, maintenance and repair with emphasis on remote-handling techniques. This paper will focus on elements 2 and 4 above and, in keeping with the conference participation from both fusion and fission programs, will emphasize potential interfaces between fusion technology and experience in the fission industry

A comparison of stresses in armor joints with and without interlayers

Reliable joining of armor to heat sinks for plasma facing components has been a persistent problem in fusion and a concern for the International Thermonuclear Experimental Reactor (ITER). Post-fabrication and operating stresses in heat sinks with a 1mm compliant layer (or no interlayer) between tungsten armor and a CuCrZr channel were analyzed with a 2-D finite element model with temperature dependent properties, generalized plane strain, and strain hardening

Liquid surfaces for fusion plasma facing components—A critical review. Part I: Physics and PSI

This review of the potential of robust plasma facing components (PFCs) with liquid surfaces for applications in future D/T fusion device summarizes the critical issues for liquid surfaces and research being done worldwide in confinement facilities, and supporting R&D in plasma surface interactions. In the paper are a set of questions and related criteria by which we will judge the progress and readiness of liquid surface PFCs. Part-II (separate paper) will cover R&D on the technology-oriented aspects of liquid surfaces including the liquid surfaces as integrated first walls in tritium breeding blankets, tritium retention and recovery, and safety

Thermal-hydraulic/thermo-mechanical testing of solid breeder blanket modules using microwave energy

This report provides the design and analyses leading to experimentation with a solid-breeder blanket module utilizing microwave energy as the volumetric heat source. (MOW

In situ spectroscopic measurements of erosion behavior of TFTR-redeposited carbon materials under high-flux plasma bombardment in PISCES-A

The chemical erosion behavior of graphite materials pre-exposed in Tokamak Fusion Test Reactor (TFTR) as the bumper limiter has been investigated spectroscopically under deuterium plasma bombardment in the PISCES-A facility. The deuterium plasma bombardment conditions are: ion bombarding energy of 300 eV; ion flux of 1.7 /times/ 10/sup 18/ ions s/sup /minus/1/ cm/sup /minus/2/; plasma density of 1.4 /times/ 10/sup 12/ cm/sup /minus/3/; electron temperature of 11 eV; and neutral pressure of 3 /times/ 10/sup /minus/4/ torr. The chemical erosion yield is measured with CD-band spectroscopy during the temperature ramp from 100 to 900/degree/C at an average rate of about 5 degrees/s. The materials used include virgin POCO AFX-5Q graphite, graphite tile pieces from the redeposition-dominated and erosion-dominated areas of the bumper limiter in TFTR. It has been found in common for these graphite materials that the chemical erosion yield maximizes at a temperature around 550/degree/C. However, graphite from the redeposited area has shown a somewhat higher maximum erosion yield and significantly steeper temperature dependence. In addition, the removability of the redeposited materials by helium plasma bombardment has been studied. The removal rate is found to be similar to the physical sputtering yield of carbon by helium. The surface morphology and surface composition has been analyzed with SEM and EMPA in parallel with these erosion behavior measurements. 38 refs., 5 figs., 1 tab

Net plasma facing materials behaviour under erosion and redeposition regimes in PISCES

Submitted to Nuclear FusionSIGLEITItal