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The amount of tritium remaining within and the release rate out of stainless steel process waste from the 232-F Tritium Facility at SRS is calculated as a function of time using the historical exposure of pipe during operation of the facility (1955-1958) and its subsequent deactivation and lay-up. The solution and diffusion of tritium in the wall is the mechanism that governs both the tritium contamination of the pipe during operation and its gradual release after deactivation, including radioactive decay while in the metal. This analysis applies to Segment 6 of the so-called Components in Grout waste form. Results of these calculations will be used in the Groundwater Transport assessment, part of the analysis of the Components in Grout

CALCULATION OF TRITIUM RETENTION AND RELEASE FROM COMPONENTS IN GROUT- SEGMENT 6 METALLIC WASTE FROM DEMOLISHED BUILDING 232-F