Abstract: An alluminum-clad low-enrichment, uranium-alloy fuel element of flat plate configuration has been proposed for the Organic Moderated Reactor (OMR).

Abstract: The data presented herein, in the form of graphs, can be used to obtain the value of this energy.

Abstract: This report presents the results of plant optimization studies and cost estimates of the reference design for a natural uranium, graphite moderated, gas-cooled reactor and power plant which was described in NAA-SR-1833.

Abstract: Presented herein is the preliminary design of a natural uranium, graphite moderated, CO2-cooled reactor and power plant similar to, but larger than, the British Calder Hall plant, with a net electrical output of 130 MWE.

A preliminary study has been made of a solid homogeneous reactor for stationary power plant application. The core consists of graphite spheres impregnated with uranium and thorium, and the coolant is bismuth. This concept possible offers advantages over other solid fuel reactor systems with respect to simplification of core structure, fuel fabrication and fuel handling, and reduction of fuel inventory external to the reactor. From the results of this preliminary study, it appears that the potential cost of electric power from this reactor is competitive with that from other reactor systems which have been proposed for the same application. The Po210 produced in the coolant presents a decontamination problem, but is also possibly a valuable by-producgt.

Purpose: Automatic recording equipment was designed for use with a high temperature Sykes experiment in which calorimetric measurements were to be made to temperatures approaching 2000* C. At such high temperatures, radiation becomes the dominant mechanism for heat transfer. The temperature differences which are used to determine the magnitude of this transfer no longer are directly proportional to it, but must be related by the Stefan-Boltzman law of radiation.

"General Chemistry investigations reported herein includes: (1) the Organic Coolant-Moderator Program, (2) investigations on zirconium hydride, and (3) analytical chemistry."

Abstract: In the present study, the variation of moderator composition was postulated to be effected by the inclusion of varying amounts of beryllium oxide in the graphite of an SGR.

Design and operational techniques are described for a freeze seal type centifugal pump for use in the Sodium Reactor Experiment.

Abstract: A calculational method is presented which may be used to determine fast and thermal neutron flux distributions at deep neutron penetrations in hydrogenous shields.

"An experiment to determine the effect of reactor irradiation on the thermal conductivity of uranium-impregnated graphite at elevated temperatures as described. The results show a decrease in the thermal conductivity saturating at [approximately] 60 percent at a temperature of 700 degrees C; at [approximately] 50 percent at a temperature of 1000 degrees C; and at [approximately] 25 percent at a temperature of 1300 degrees C. It was found that after irradiation at a given temperature, exposure at a higher temperature resulted in an increase in the thermal conductivity. The converse was also observed. Within the precision of measurement there was no difference in effed between temperature changes produced by varying the fission rate in the samples and changes produced by varying the power in an external heater."

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Abstract: The fast and thermal neutron leakage from the 30 megawatt SGR-P4 reactor has been studied by three independent methods.

The design of a waste-gas treatment system for organic moderated reactors requires a knowledge of reactor waste-gas composition, generation rate, and radioactivity. To obtain data on these variables a continuous stream analyzer was constructed to analyze the waste gas from the Organic Moderated Reactor Experiment (OMRE).

Abstract: This report presents a reference design of an organic moderated and cooled reactor for the propulsion of a 38,000-dwt tanker.

Introduction: This memo gives a general description of the components and equipment affecting the control of the SRE, and the equipment associated with all reactor services.

"Heat production resulting from the absorption of gamma ray photons in thermal shields and the leakage of neutrons and photons from ferritic thermal shields are investigated. The gamma rays considered arise from three types of reactor radiation -- thermal neutrons, fast neutrons, and core and reflector gammas. The energy spectra of the fast neutron leakage and absorption have been investigated in some detail because of the significant contribution of fast neutrons to the heating of the concrete biological shield."

From abstract: A conceptual design study has been performed for a sodium cooled, graphite moderated, thermal power-breeder reactor utilizing the Thorium-Uranium 233 breeding cycle. Several aspects of the design of the system are considered but no attempt has been made to supply all the details. It appears that the design presented is feasible and will allow the production of economic power as well as full utilization of thorium resources.

Abstract: A magnetic-jack control-safety rod is under development for the 45.5 thermal megawatt Organic Moderated Reactor. The rod is "unitized," i.e., the poison element, drive, position indicator, and shock absorber are contained in a compact assembly which is inserted in a regular fuel channel opening in the core. Tests to develop components capable of operating under these conditions are described and results are reported.

"The composition of the solder for the solder plug has been set as the tin-silver eutectic. Final tests on this solder show that life expectancies much longer than 6 months are probable with the current design. The design of the heater tube to contain the solder plug has been settled. This consists of a copper tube impregnated with U235O2. Arrangements have been made to have test specimens fabricated by powder metallurgy techniques. The equipment for the MTR in-pile test of trigger element response times has been largely completed and tested. The design of the complete inner capsule for the BF3 safety element has been developed as well as the cladding technique. Mock-up elements have been tested in the Hanford test reactor to determine the control that may be obtained with elements of this type, although the analysis of the results has not been made. Prototype elements are also ready for testing in the test pile, except for loading with B10F3. Experiments have been designed and submitted for approval for production pile tests of prototype."

From abstract: This paper presents temperature-entropy and Mollier charts for sodium, and describes briefly the method used for their construction, based upon data from the literature.

"Engineering pertinent to the development of the sodium-cooled, graphite-moderated type of reactor was continued. This included work on problems related to the zirconium canned moderator, low enrichment uranium fuel, sodium piping, secondary coolant system, shielding, and the control and safety elements. A large fraction of the work was devoted specifically to problems of the proposed Sodium Reactor Experiment (SRE) configuration. In this connection, an integrated effort was initiated to prepare a complete preliminary design of the SRE by an early date. In addition, two alternate sodium-graphite reactor configurations were studied. One was an intermediate size, 145 thermal megawatt, unit optimized for the production of low cost plutonium. The second was a low power 10 thermal megawatt intended for power production, but in which sodium circulation through the core was entirely dependent upon thermal convection."