UNCLASSIFIED RESEARCH AND DEVELOPMENT PROGRAMS EXECUTED FOR THE DIVISION OF REACTOR DEVELOPMENT AND THE DIVISION OF RESEARCH, JULY 1960

Technical Report · Wed Aug 10 00:00:00 EDT 1960

OSTI ID:4060751

McEwan, L.H. comp.

Plutonium Recycle Program. Extrusion billets of PuAl- Ni alloy were cast and exhibited satisfactory corrosion resistance. Metallographic examinations of Al-Pu rods indicated that very few, if any, microstructural changes occurred in the core due to irradiation. Internal pressures produced at elevated temperatures by gases desorbed from UO/sub 2/ contained in PRTR Mark I fuel elements were measured. The applicability of the Magnetic Force Butt Welding Closure Process to materials other than stainless steel and zircaloy was investigated. The erosion effect of stainless steel on zircaloy-2 was determined as a function of temperature in the modified stirring autoclave. The overall PRTR Project was estimated to be about 97.5% complete. Experiments were performed to determine the attack by zirflex decladding solutions on Al-Pu-Ni- Si alloy spike fuel cores. Kinetics of the adsorption of thorium nitrate complex ions on Permutit SK anion exchanger were studied. An electrolysis cell sized to permit production of pound lots of UO/sub 2/ was placed in operation. Study of nonaqueous separation processes showed that there appear to be economic prospects for those processes from which uranium of less than natural enrichment can be discarded. Plutonium Ceramics Research. Preliminary investigation of the PuO/ sub 2/- ZrO/sub 2/ phase diagram showed phase boundaries at room temperature equilibrium to be at approximately 40 and 70 wt.% PuC/sub 2/. Uranium Dioxide Fuels Research. Analyses of gases released from as-received, fused UO/sub 2/ during vacuum annealing at 800 deg C indicated approximately 1 ppm of hydrogen and argon. Measurements of thermal conductivity of irradiated UO/sub 2/ were resumed. In-Reactor Measurements of Mechanical Properties. Data were accumulated on the in-reactor creep rates of a zircaloy-2 specimen. Activation energies for creep were measured on a series of cold-worked zircaloy-2 specimens at various test temperatures and stress levels. Gas Cooled Power Reactor Program. Graphite did not react with high-pressure CO/sub 2/ in the temperature range 360- 660 deg C during exreactor tests for as long as 298 days. Helium density measurements to determine graphite grain densities are being made to investigate the possibility that the ultimate contraction of graphite may not be limited by the theoretical density of graphite. Nondestructive Testing Research. A study of the use of orthogonalized exponentials in signal analysis was continued. Radioactive Residue Processing Development. An experiment was performed in which diluted high-level Purex waste was passed through three short columns of clinoptilolite in series. Results indicated a cesium capacity of 27 and 35 bed volumes of undiluted waste. Radiation Effects on Metals. Radiation damage recovery is being studied for a number of metals. (M.C.G.)

• Research Organization: General Electric Co. Hanford Atomic Products Operation, Richland, Wash.
• DOE Contract Number: AT-(45-1)-1350
• NSA Number: NSA-15-014333
• OSTI ID: 4060751
• Report Number(s): HW-66448
• Resource Relation: Other Information: Orig. Receipt Date: 31-DEC-61
• Country of Publication: United States
• Language: English

Related Subjects

CHEMISTRY
ACTIVATION ENERGY
ADSORPTION
ALUMINUM
ANNEALING
ARGON
CARBON DIOXIDE
CASTING
CERAMICS
CESIUM
CLINOPTILOLITE
COMPLEXES
COOLING
CORROSION
CREEP
DENSITY
ELECTROLYTIC CELLS
ENRICHMENT
EXTRUSION
FUEL ELEMENTS
FUELS
GASES
GCPR
GRAPHITE
HELIUM
HYDROGEN
IONS
IRRADIATION
MATERIALS TESTING
MEASURED VALUES
MECHANICAL PROPERTIES
METALLOGRAPHY
METALS
NICKEL
NONDESTRUCTIVE TESTING
PERMUTIT
PHASE DIAGRAMS
PLUTONIUM
PLUTONIUM ALLOYS
PLUTONIUM OXIDES
PRESSURE
PRODUCTION
PUREX PROCESS
RADIATION EFFECTS
REACTION KINETICS
REACTORS
RESEARCH REACTORS
SEPARATION PROCESSES
SILICON
STAINLESS STEELS
TEMPERATURE
TESTING