Tritium retention and release analysis for US-ITER blanket [electronic resource].
- Washington, D.C. : United States. Dept. of Energy, 1990.
Oak Ridge, Tenn. : Distributed by the Office of Scientific and Technical Information, U.S. Dept. of Energy.
- Physical Description:
- Pages: (9 pages) : digital, PDF file
- Additional Creators:
- Argonne National Laboratory
United States. Department of Energy
United States. Department of Energy. Office of Scientific and Technical Information
- The US design for the ITER tritium-breeding blanket consists of layers of Be multiplier, stainless steel cladding, and Li₂O ceramic breeder. Tritium is recovered from the ceramic breeder by purging it with He + 0.2% H₂. Models have been developed to describe the purge-flow thermal-hydraulics and gas reactions and the tritium retention/release due to lattice diffusion, desorption/adsorption, solubility/precipitation, and percolation through interconnected porosity. These have been incorporated into the steady-state code TIARA for the purpose of performing design calculations for Tritium Inventory and Release Analysis. Transient calculations for pulsed operation are done with a modified version of the DISPL code. The results of both steady-state and transient analyses for tritium retention and releases are given for anticipated ITER operating conditions. 13 refs., 6 figs., 3 tabs.
- Published through SciTech Connect.
9. topical meeting on technology of fusion energy, Oak Brook, IL (USA), 7-11 Oct 1990.
Lin, C.C.; Billone, M.C.; Gohar, Y.; Attaya, H.
- Funding Information:
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