Tritium control in helium-cooled blankets
Abstract
As a part of the Blanket Comparison and Selection study (BCSS), GA Technologies as responsible for the design of helium-cooled, solid- and liquid-metal breeder blankets. Conceptual blanket designs are developed, including the consideration of the generation, transport, and extraction of tritium. Evaluations are made of the inventory and leakage of tritium for helium-cooled Li2O and LiAlO2 and liquid lithium breeder blankets for tokamak and tandem mirror reactors. To facilitate the evaluation, a solid breeder tritium code TRIT4 is developed. The results from this study indicate that tritium inventories and leakages are acceptable for the proposed helium-cooled blankets. An assumption made in the tritium leakage calculations is that tritium is released to the helium purge and coolant streams as T2 and remains in that form. If oxidation to T2O is possible, significant reduction in the tritium leakage will be possible. It is concluded that more experimental data on breeder material properties and tritium permeation behavior are needed. However, an adequate number of techniques is available to control the breeder tritium inventory and leakage to an acceptable level in helium-cooled solid- and lithium-breeder blankets.
- Publication:
-
Presented at 2nd Natl. Topical Meeting on Tritium Technol. in Fission
- Pub Date:
- June 1985
- Bibcode:
- 1985ttff.meet.....W
- Keywords:
-
- Aluminum Oxides;
- Blankets (Fission Reactors);
- Breeder Reactors;
- Ceramics;
- Controllers;
- Coolants;
- Helium;
- Isotopic Enrichment;
- Lithium Oxides;
- Oxidation;
- Tritium;
- Inventories;
- Liquid Metals;
- Performance Tests;
- Tandem Mirrors;
- Nuclear and High-Energy Physics