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Improved Method for Numerically Solving Multi-Group Reactor Equations

Description: "A method for solving multi-group reactor equations which arise in the diffusion approximation is outlined. Considerable work has been done on this problem at KAPL and ORNL. Their approach is to replace the differential equations by difference equations. Complications arise in this method where more than one slowing down medium is present since the fluxes are discontinuous at the interfaces. The primary purpose of this article is to develop an exact integral expression for the neutron flux w… more
Date: September 15, 1954
Creator: Lehman, G. W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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The Distribution of Tracer Plutonium and Fission Products Between Molten Uranium and Solid Uranium Oxide, Carbide, and Nitride

Description: "A study has been made of the distribution of tracer fission products and plutonium between small samples of molten uranium and solid uranium oxide, carbine, and nitride. The distribution showed the same behavior i general for all three materials: 1. The rare earth elements, Cs, Ba, and Sr were extracted primarily into the solid scrub phase. 2. Zirconium and Nb partially concentrated in the scrub phase. 3. Plutonium, Mo, and Ru tended to remain completely in the metal phase. The distribution… more
Date: September 15, 1954
Creator: Keneshea, F. J.; Saul, A. M. & Young, C. Y.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

A Sodium Cooled, Graphite Moderated, Low Enrichment Uranium Reactor for the Production of Useful Power

Description: "A design study is presented for a sodium cooked, graphite moderated power reactor utilizing low enrichment uranium fuel. The design is characterized by dependence on existing technology and the use of standard, or nearly standard, components. The reactor has a nominal rating of 167 thermal megawatts, and a plant comprising three such reactors for a total output of 500 thermal megawatts is described. Sodium in a secondary, non-radioactive, circulation system carries the heat to a steam genera… more
Date: September 15, 1954
Creator: Weisner, Edward F.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Carbide Fuels in Fast Reactors

Description: Abstract: Cladding and fuel material processing prospects are reviewed, and fuel system possibilities for near-term (~1 mill/kwh) and long-range (<0.5 mil/kmh) fuel cycles are described.
Date: September 15, 1965
Creator: Wheelock, C. W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

OMR Control-Safety Rod Component Development Tests

Description: 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.
Date: September 15, 1959
Creator: Howell, J. D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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