Actions for SODIUM-COOLED REACTORS PROGRAM. FAST CERAMIC REACTOR DEVELOPMENT PROGRAM. Second Quarterly Report, January-March 1962

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SODIUM-COOLED REACTORS PROGRAM. FAST CERAMIC REACTOR DEVELOPMENT PROGRAM. Second Quarterly Report, January-March 1962

Author
Leitz, F. J. ed.
Published
United States : [publisher not identified], 1963.
[Oak Ridge, Tennessee] : [U.S. Atomic Energy Commission], 1963.
Physical Description
microfiche : negative ; 8 x 13 cm

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Summary
Progress is reported on the development of the Experimental Fast Ceramic Reactor and Fast Ceramic Reactor and demonstrations of low fuel cycle cost capability for the reactors. Further design work was accomplished on a fast reactivity excursion device for the EFCR. Preliminary experiments were accomplished involving gas charging used for pneumatic activation. Modification of the conceptual design of the original EFCR facility and scoping of the EFCR for a broader experimental program than originally proposed is reported. Evaluation of the transport of fission products in previously irradiated Phase I specimens was accomplished. Transient testing of EFCR-type fuel rods is reported. A study comparing the economic potential of mixed carbide fuel with mixed oxide fuel for fast reactors was conducted. A study was made of the processes available for fabrication of plutonium-containing fuel for a fast ceramic reactor, and for chemical reprocessing of irradiated fuel. A 37 fuel pin FCR prototype assembly and a 19 (individually encapsulated) FCR fuel pin assembly were designed for prospective fast flux testing in the EBR-II reactor. A single fuel pin test capsule was designed for the Dounreay Fast Reactor. Refinement and debugging of the FORE digital computer code, designed to analyze tue transient thermal-nuclear behavior of fast reactors, has proceeded to the point where the code is being run on a routine basis. The transient characteristics of a large fast ceramic reactor configuration were investigated using the FORE code. The preliminary design status of a 500 Mw(e) FCR power plant is summarized. The EICA (Elastic -Inelastic Cross Section DELTA veraging) computer code is specified in sufficiert detail to permit coding to start. The RAPTURE computer code designed for the computation of Doppler coefficients in fast reactors is described. Technical evaluation of a proposed series of FCR critical experiments to be carried out in the ZPR III facility was made. (N.W.R.)
Report Numbers
GEAP-3957
Other Subject(s)
Collection
U.S. Atomic Energy Commission depository collection.
Note
DOE contract number: AT(04-3)-189
NSA number: NSA-17-031678
OSTI Identifier 4676012
Research organization: General Electric Co. Atomic Power Equipment Dept., San Jose, Calif.
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