Abstract
A specimen of Li 7Pb 2 (lithium in natural abundance) contained in a Zircaloy capsule was irradiated in the Oak Ridge Research Reactor Poolside Facility for ~ 700 h. The design of the irradiation assembly (designated TBC-07) permitted operational control of the breeder material at a temperature of (390 ± 25)°C throughout the irradiation. The amount of tritium determined to be present in the capsule parts during post-irradiation chemical analysis was (52 ± 5)Ci, corresponding to the burnup of ~ 8% of the original 6Li in the capsule. The estimated tritium production based on simple neutronics calculations, using an assumed total flux of 1.2 × 10 14 neutrons/cm 2 s, was ~56 Ci. The value derived from a limited dosimetry measurement on the TBC-07 thermocouple wire was ~ 80 Ci. The axial tritium distribution profile along the Zircaloy capsule wall and the Li 7Pb 2 cylinder indicated a nonuniform temperature during irradiation, with the capsule ends probably being at lower temperature than the center region. There was no evidence of tritium loss from the experiment during or after irradiation. Metallurgical analysis of the Li 7Pb 2 after irradiation revealed the presence of a unique network of hexagonal bubbles wherein facets are preferentially aligned along crystallographic planes and are also rich in lithium.
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