This report summarizes work performed up to June 1973 relative to the description of the behavior of tritium in a high-temperature gas-cooled reactor; the main objective was the development of a computer code with which accurate assessment of tritium distribution in HTGRs could be made. The resulting code, TRITGO, is operational and, subject to the limitations discussed below, may be used for determining parameter sensitivity and for defining future research objectives. However, many extensions and amplifications that were evidently desirable were not included. Some of these are indicated in the text. (For example, the special role of the fuel element purge system in removing tritium released by the fuel in a reactor such as the Peach Bottom HTGR is not fully described.) In addition, it is now recognized that a slow, temperature-dependent release of tritium must occur from fuel particles, as well as from graphite and other solids. In the present version of TRITGO, this is handled by a partition factor corresponding to immediate release of a small fraction and permanent retention of the remainder. Similarly, fuel replacement in the core is not presently incorporated in the illustrative calculation. Representative values of the input parameters have been taken from PSAR descriptions of the Fort St. Vrain Reactor. The relationship between code output and the performance of an HTGR remains to be established critically.
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