Assessment of erosion and surface tritium inventory issues for the ITER divertor

J. N. Brooks, R. Causey, G. Federici, D. N. Ruzic

Research output: Contribution to journalArticlepeer-review

Abstract

We analyzed sputtering erosion and tritium codeposition for the ITER vertical target divertor design using erosion and plasma codes (WBC/REDEP/DEGAS+) coupled to available materials data. Computations were made for a beryllium, carbon, and tungsten coated divertor plate, and for three edge plasma regimes. New data on tritium codeposition in beryllium was obtained with the tritium plasma experiment (TPE) facility. This shows codeposited H/Be ratios of the order of 10% for surface temperatures ≤300°C, beryllium thereby being similar to carbon in this respect. Hydrocarbon transport calculations show significant loss (10-20%) of chemically sputtered carbon for detached conditions (Te ≈ I eV at the divertor), compared to essentially no loss (100% redeposition) for higher temperature plasmas. Calculations also show a high, non-thermal, D-T molecular flux for detached conditions. Tritium codeposition rates for carbon are very high for detached conditions (∼ 20 g T/1000 s discharge), due to buildup of chemically sputtered carbon on relatively cold surfaces of the divertor cassette. Codeposition is lower (∼ 10X) for higher edge temperatures (∼ 8-30 eV) and is primarily due to divertor plate buildup of physically sputtered carbon. Peak net erosion rates for carbon are of the order of 30 cm/burn yr. Erosion and codeposition rates for beryllium are much lower than for carbon at detached conditions, but are similar to carbon for the higher temperatures. Both erosion and tritium codeposition are essentially nil for tungsten for the regimes studied.

Original languageEnglish (US)
Pages (from-to)294-298
Number of pages5
JournalJournal of Nuclear Materials
Volume241-243
DOIs
StatePublished - Feb 11 1997

Keywords

  • Erosion and particle deposition
  • High Z wall material
  • ITER
  • Low Z wall material
  • Tritium inventory

ASJC Scopus subject areas

  • Nuclear and High Energy Physics
  • General Materials Science
  • Nuclear Energy and Engineering

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