The fundamental objective of the graphite temperature limit is to prevent excessive oxidation of the graphite moderator blocks with carbon dioxide and water vapor in the reactor atmosphere. Laboratory tests have shown that 10 percent uniform oxidation of graphite results in a loss in strength of approximately 50 per cent. This value of a strength loss of 50 percent has been arbitrarily established as the limit to graphite oxidation over the lifetime of the reactor stack. This limit has been further arbitrarily interpreted as an oxidation rate of not more than two percent per thousand days of reactor operation. At this rate limit, the reactor stack could be operated for 15--20 years before the graphite strength was reduced 50 percent. As with any chemical reaction, the rate of reaction of graphite with oxygen, carbon dioxide, and water vapor increases with increasing temperature. The graphite temperature limits presently in effect at the smaller reactors were therefore primarily established to hold the reaction rate between graphite and carbon dioxide at two percent per thousand operating days and were based on laboratory and in-reactor experiments which are presently believed to have been partially contaminated with oxygen from air leakages. It is therefore the …

The 02 Research and Development program for the Irradiation Processing Department is contained in seven 189-forms, R-1 through R-6 and M-1 and M-2. The subprograms designated by R pertain to reactor research and development, and those designated by M pertain to the metallurgy or the fuels research and development. This report covers the progress in each of the subprograms on an individual basis and follows the format proposed by RLOO. Although the reports are brief, the results are presented in a semitabular form and are essentially complete. Only the activities of the first four months of the fiscal year are included. It should be noted that the over-all IPD program is based on a nearly uniform expenditure of funds over the entire year and the expenditures to date reflect this general pattern. A number of items are listed in the reports on the subprograms that have been performed on product cost funds but were made possible by the research and development effort.

The amount of enriched uranium to support various type of target elements was determined in the Hanford Test Reactor. This report describes the measurement of separate and striped loadings for canned thorium oxide, bismuth, lithium-aluminum, and, in addition, bare depleted uranium. This experiment was undertaken to provide information needed to estimate the E-metal requirements in a production reactor when replacing natural uranium with target material. The amount of enriched uranium which is required to maintain the reactivity in support of target materials must be known to properly predict total reactor reactivity. Danger coefficient measurements were made during original startup tests to provide this information. However, fuel dimensions have changed and new target materials have been introduced which have not been tested in this manner. The target of particular interest at this time of thorium oxide.

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This test request defines pre-test data for the irradiation of enriched yttrium-uranium hydride samples in the LITR C-48 facility.

Introduction: This report has been prepared to document the extensive analytical work required to design the control system for the Sodium Components Test Installation (SCTI), constructed for the Atomic Energy Commission at the Atomics International field laboratory.

Abstract: One hundred gram quantities of uranium dioxide, irradiated to burnup as great as 21,000 Mwd/MTU and previously reprocessed by AIROX (Atomics International Reduction Oxidation), were refabricated into high density pellets and encapsulated for re-irradiation.

Introduction: A series of critical and subcritical experiments have been carried out as part of the Advanced Sodium Graphite Reactor (ASGR) Critical Experiments program to provide physics information which would be directly applicable to the AGSR concept.

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Declassified 31 Aug 1973. The PWAR-20 reactor and shield are being developed for the SNAP-50/SPUR power plant (300-kW(e), 10,000 h). Clad UC and UN were irradiated; swelling and gas release were measured. Nb-1 Zr, D-43, and T- 111 were also evaluated. (DLC)

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