SODIUM COOLED REACTORS PROGRAM--FAST CERAMIC REACTOR DEVELOPMENT PROGRAM FOURTH QUARTERLY REPORT, JULY-SEPTEMBER 1962

Technical Report · Thu Oct 31 00:00:00 EST 1963

OSTI ID:4728305

Leitz, F.J. ed.

The fifth sodium logging capsule was irradiated under the severest test conditions to-date, yet cladding damage did not occur. Experimental design of the Series II capsules is underway and mixed oxide fuel specimens are being fabricated. The FCR sodium loop was operated successfully at design conditions for over 400 hours. Fabrication of hyperstoichiometric uranium dioxide fuel was started. Conceptual design of a compact dynamic sodium capsule was completed. Programmed examination of Series I large diameter UO/sub 2/ fuel specimens was completed. Five capsulea were irradiated in TREAT and examined to identify incipient failure limits for single and repeated transients. Capsule design for Series II was completed and test assemblies are in the final stages of fabrication. The hazard analysis and test specifications for series II and III were submitted to the manager of the TREAT facility, and testing will proceed as soon as hazards approval is received. Several batches of mixed oxide pellets satisfying density and dimensional specifications were produced for use in the FCR fuel testing program. Equipment is being set up for the purpose of determining the oxygen to heavy metal ratio in mixed oxide fuel by means of the thermogravimetric technique. Scouting studies showed that the plutonium hydroxide-ammonium diuranate precipitate can be directly reduced and sintered to a high density compactable grade oxide in one firing at approximately 1500 C. Post-irradiation examination of high exposure mixed oxide fuel ()70,000 Mwd/t) operated at approximately 20 kw/ft did not reveal gross plutonium migration; the migration observed was within the limits of error, approximately 10%. Earlier physics calculations on fuel cycles and reactivity coefficients for the FCR reference design were repeated using updated methods and cross sections. The newer cross sections and calculation methods produced an improvement of the fuel cycle economics and a reduction of the Doppler coefficient for the reference core size and composition. Calculations were made to evaluate the effect of composition homogeneity in FCR on the effective absorption cross sections of U/ sup 238/ and Pu/sup 239/. Results indicated that the assumption of a homogeneous core is valid with insignificant error. ZPR-III experimental data on an oxide- fueled assembly were compared with multi-group diffusion calculations using the MISY code and a 21-group cross section sel, and SNG transport results show good agreement. Transient calculations using the FORE code indicated that rapid reactivity insertions of over a dollar in EFCR can be adequately simulated by a relatively slow addition using the main control system followed by rapid insertion of the remaining reactivity using the fast reactivity excursion device FORE calculations on the effects of variations in fuel-gap heat transfer coefficients and fuel thermal conductivities is to lessen the severity of an excursion for a given steady-state power and a given reactivity input. (auth)

• Research Organization: General Electric Co. Atomic Power Equipment Dept., San Jose, Calif.
• DOE Contract Number: AT(04-3)-189
• NSA Number: NSA-17-015390
• OSTI ID: 4728305
• Report Number(s): GEAP-4080
• Resource Relation: Other Information: Orig. Receipt Date: 31-DEC-63
• Country of Publication: United States
• Language: English

Related Subjects

REACTOR TECHNOLOGY
BURNUP
CAPSULES
CERAMICS
COMPUTERS
CONTROL SYSTEMS
CROSS SECTIONS
DENSITY
DIFFERENTIAL EQUATIONS
DIFFUSION
ECONOMICS
EQUATIONS
EXCURSIONS
FABRICATION
FAILURES
FUEL CANS
GROUP THEORY
HEAT TRANSFER
HEAT TREATMENTS
HOMOGENEOUS REACTORS
IRRADIATION
LABORATORY EQUIPMENT
LIQUID METAL COOLANT
OXYGEN
PELLETS
PLANNING
PLUTONIUM 239
PLUTONIUM COMPOUNDS
PRECIPITATION
PROGRAMMING
QUANTITY RATIO
REACTIVITY
REACTOR SAFETY
REACTORS
RESEARCH REACTORS
RESONANCE ESCAPE PROBABILITY
SINTERING
SPECTRAL SHIFT
STANDARDS
THERMAL CONDUCTIVITY
TRANSIENTS
TRANSPORT THEORY
TREAT
URANIUM 238
URANIUM COMPOUNDS
URANIUM DIOXIDE
U