Sodium Graphite Reactor Quarterly Progress Report for October-December 1955. Section A. Section B

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An analysis was made of the nuclear parameters for sodium graphite reactor lattices. These parameters include thermal utilization, macroscopic cross sections, thermal diffusion length, and neutron absorption. Results of all calculations are given in graphical form. Test fuel slugs for the SRE were cycled up to 500 times between 100 and 500 deg C at the rate of 2 cycles/hr. Results are tabulated. The centrifugal casting of U alloy fuel slugs is briefly evaluated. Results of the microscopic examination of the extruded ThU breeder fuels are shown. The percent elongation of graphite due to the presence of Na is shown ... continued below

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Pages: 100

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Martin, A. B. & Cochran, J. C. April 15, 1956.

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An analysis was made of the nuclear parameters for sodium graphite reactor lattices. These parameters include thermal utilization, macroscopic cross sections, thermal diffusion length, and neutron absorption. Results of all calculations are given in graphical form. Test fuel slugs for the SRE were cycled up to 500 times between 100 and 500 deg C at the rate of 2 cycles/hr. Results are tabulated. The centrifugal casting of U alloy fuel slugs is briefly evaluated. Results of the microscopic examination of the extruded ThU breeder fuels are shown. The percent elongation of graphite due to the presence of Na is shown for various temperatures. Results of wear tests on graphite are also tabulated. The behavior of Zr in liquid Na was studied, and weight gains in Zr are summarized. Analog computer studies were continued, and data are included on the temperature effects of the response time of coolant channel Na outlet temperature thermocouples, the effects of continuous rod motion and pump speed changes on the outlet Na temperature and power, and the outlet temperature as a function of scram time. The critical evaluation of B--Ni rods is tabulated. The fuel rod assembly apparatus is described. Fuel rod development is discussed. Cyclograph traces of rods bonded with various Na--K alloys were recorded for rods at room temperature and heated to 450 and 600 deg F. The traces are indicative of uniform bonding. The moderator can fabrication and testing is also discussed. Tests were completed on Freeze Seal No. 2 for the 6-in. oval port Wedgeplug test valve at 450, 850, and 1250 deg F. The temperature gradients from the hot flange face to the end of the seal mechanism for various valve temperature conditions are shown. Sodium leak rates through the valve are tabulated. Progress in the development of a liquid Na level gage is briefly reported. The tubular heater experiment was completed, and the times to raise pipe temperatures from ambient to 350 deg F are tabulated. Designs for a 6-in. Na pump loop are described briefly. A one to 3.5 scale model of a SRE fuel element was constructed to study the effect of side drag on the element during insertion operations at those fuel channels located near the outlet of the upper plenum chamber. The calibration of the SRE fuel element orifices was studied. Control rod lead screw development is discussed. Development of the safety rod system is described. Core tank galling tests are summarized. Experiments to determine the effects of radiation on MoS/ sub 2/ are described. Dose buildup factors for the concretes to be used in the reactor top shield are tabulated. Constants for the quadratic representation of the dose buildup factor and the capture gamma rays from heavy concrete are also tabulated. The sodium pumps and service system are described. The rated cooling capacity of the SRE tetralin evaporative cooler was checked. Results of a study on the effects of NaK temperature on the H content of He are tabulated. (D.E.B.)

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Pages: 100

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  • Other Information: Decl. Apr. 8, 1957. Orig. Receipt Date: 31-DEC-58

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  • Report No.: NAA-SR-1582
  • Grant Number: AT(04-3)-49
  • DOI: 10.2172/4329393 | External Link
  • Office of Scientific & Technical Information Report Number: 4329393
  • Archival Resource Key: ark:/67531/metadc1023765

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  • April 15, 1956

Added to The UNT Digital Library

  • Oct. 15, 2017, 10:09 p.m.

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  • Jan. 22, 2018, 11:26 a.m.

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Martin, A. B. & Cochran, J. C. Sodium Graphite Reactor Quarterly Progress Report for October-December 1955. Section A. Section B, report, April 15, 1956; Canoga Park, California. (digital.library.unt.edu/ark:/67531/metadc1023765/: accessed September 24, 2018), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.