Release Characteristics of Tritium from High-Purity Lithium Oxide
- 1 September 1985
- journal article
- research article
- Published by Taylor & Francis in Fusion Technology
- Vol. 8 (2P2) , 2054-2060
- https://doi.org/10.13182/fst85-a24587
Abstract
Rates of tritium release from neutron-irradiated lithium oxide were determined from isothermal release experiments. High-purity, monocrystalline lithium oxide was purged ex-reactor with helium and helium-hydrogen gas streams. Overall release was found to be controlled by solid-phase diffusion, and was predominantly in the form of condensible species. The diffusion coefficient, D, was given by The result of an independent concentration profile analysis at 923 K was in agreement with the gas release diffusion coefficient. Sweeping the Li2O with hydrogen-containing gas was found to enhance tritium removal during the early stage of each run.Keywords
This publication has 11 references indexed in Scilit:
- Measurements of the activity coefficient of LiOH dissolved in Li20(s) for an evaluation of Li2O as a tritium breeder materialJournal of Nuclear Materials, 1984
- In-situ tritium recovery experiment from lithium oxide under high neutron fluenceJournal of Nuclear Materials, 1984
- Preparation, Characterization, and Melting Point of High‐Purity Lithium OxideJournal of the American Ceramic Society, 1983
- Diffusion of tritium in single crystal Li2OJournal of Nuclear Materials, 1983
- Thermal release of tritium produced in sintered Li2O pelletsJournal of Nuclear Materials, 1983
- A preliminary in-pile test of tritium release from Li2O pelletsJournal of Nuclear Materials, 1981
- Kinetic studies of the trttium release process in neutron-irradiated Li2O and LiOHJournal of Nuclear Materials, 1981
- The fundamental optical absorption edge and an estimation of the number of displaced atoms by the 6Li(n, α)3H reaction in Li2OJournal of Nuclear Materials, 1978
- Removal of tritium from fusion reactor blankets. Annual report, FY 1977Published by Office of Scientific and Technical Information (OSTI) ,1977
- Removal of tritium from solid CTR blanket materials. Progress reportPublished by Office of Scientific and Technical Information (OSTI) ,1975