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Using the tritium plasma experiment to evaluate ITER PFC safety

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Abstract

The Tritium Plasma Experiment was assembled at Sandia National Laboratories, Livermore and is being moved to the Tritium Systems Test Assembly facility at Los Alamos National Laboratory to investigate interactions between dense plasmas at low energies and plasma-facing component materials. This apparatus has the unique capabilty of replicating plasma conditions in a tokamak divertor with particle flux densities of 2 × 1023 ions/m2.s and a plasma temperature of about 15 eV using a plasma that includes tritium. An experimental program has been initiated using the Tritium Plasma Experiment to examine safety issues related to tritium in plasma-facing components, particularly the ITER divertor. Those issues include tritium retention and release characteristics, tritium permeation rates and transient times to coolant streams, surface modification and erosion by the plasma, the effects of thermal loads and cycling, and particulate production. An industrial consortium led by McDonnell Douglas will design and fabricate the test fixtures.

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    For example, at the Technical Meeting on Experimental Approach to the Physics of the High Density Divertor, Garching Germany, February 25–27, 1993, a system with an acquisition cost of $30M and correspondingly high operating costs was discussed. This system will require less than one-tenth that.

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Prepared for the U.S. Department of Energy, Office of Energy Research under DOE Idaho Field Office Contract DE-AC07-76ID01570.

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Longhurst, G.R., Anderl, R.A., Bartlit, J.R. et al. Using the tritium plasma experiment to evaluate ITER PFC safety. J Fusion Energ 12, 115–119 (1993). https://doi.org/10.1007/BF01059365

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Key words

  • tritium
  • implantation
  • experiments
  • safety
  • ITER