Sm-2 Reactor Core and Vessel Review Report for Period December 15, 1959 to March 18, 1960

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The SM-2 core lifetime was calculated and stuck rod criteria formulated. The control system could shutdown the core under the most adverse conditions with any single rod stuck out. The one-half-inch Eu/sub 2/O/sub 3/ flux supressors adequately supressed the power spike at the bottom of the core. The nuclear effects of increasing the dimensione of the fuel matrix were calculated. A new analysis of various reactivity measuremerts performed durirg the SM-2 critical experiments was made to test the validity of the calculational modes. Twodimensional temperature distributions for the stainless steel vessel shell and flange and for the SM-2 control rod plates were plotted. The loss of flow experiment was completed on the SM-1 at Fort Belvoir, and the results agreed with analytical predictions. The SM-2 fuel element, control rod, core support structure, and vessel designs were completed. Three steel mills indicated they could supply low cobalt, low tantalum type 347 stainless steel. Corrosion testing of irradiated boron-stainless steel was reactivated. Corrosion and impact testing of nut end bolting materials was scheduled. Dysprosium oxide was considered as an alternate absorber material. Two modified SM-2 fuel aesemblies were fabricated for testing in the Westinghouse Test Reactor loop, and two fuel elements were fabricated for insertion in the SM-1 core. Work on controlled addition of a burnable poison, such as ZrB/sub 2/, in fuel plates was continued. Procedures for welding fuel plates to side plates were completed. Automatic weld sequence timing was adopted for improvirg product quality. Pressure losses through the complete flow circuit in the reactor were recalculated and the pressure drop analysis brought up to date. Good flow distribution in the control rod fuel element was obtained in a single element rig. Methods of temperature regulation and corrosion control were established. Extended SM-2 critical experiments were initiated. Core support design modifications necessary for an accurate flow . divider mockup were completed. Effects of flux suppression upon reactivity and neutron flux at the bottom of the core were measured. Gamma dose rates were measured over the core support plate and along the control rod basket guide. Prototype testing of the control rod drive mechanism was started. The blocked channel method of measuring fuel element internal temperature in SM-2 Core I instrumental fuel assembly was selected as the reference design. (M.C.G.)