REACTOR TECHNOLOGY REPORT NO. 8. METALLURGY

Technical Report · Sun Mar 01 04:00:00 EST 1959

OSTI ID:4279336

Fuels. The formation of rare gases resulting from the fission of U and the resulting swelling of the fuel element or catastrophic volume increase are discussed mathematically. The mechanical behavior of fuel materials during reactor irradiation is discussed. A discussion fs presented on a review or irradiation effects in highly enriched fuel elements and their net engineering significance, methods of reporting performance data which are applicable to all types of solid fuel, and some performance data on Zr-U alloy fuel and the relation of these data to similar data for other highly enriched fuels. The effect of transients and longer time anneals on irradiated U-Zr alloys is reported. Several compositions within the UO/sub 2/-CeO/sub 2/ system have becn studied for possible application as ceramic fuel in nuclear reactors. Controls. The effect of temperature on the volume and structure stability of irradiated Bcontaining materials is presented. The technology and performance of Ag-In-Cd alloy control rod materials are reported. Data are presented on the attempts to improve the strength of Ag-In-Cd alloys. The fabrication of B/sub 4/C -Zircaloy- 2 dispersion samples by hot coextrusion is described. Fabrication procedures are presented for metallographic specimens contuining B/sub 4/C dispersed in Zircaloy- 2. Preliminary results of irradiation tests on B/sub 4/C, B/sub 4/C -Al/sub 2/O/ sub 3/, and B/sub 4/C -Zircaloy2 are presented. Structural Materials. The susceptibility of AISI-410 steel to stress-corrosion eracking in high temperature, high-purity water is reported. The practicality of Cr plating for corrosion protection in high-temperature, high-purity water is discussed. The cracking in type 347 stainless steel tubing has been investigated. The corrosion of Inconel by high-purity waier was investigated. The absorption of hydrogen and its effects on the properties of Zr and Zr alloys are presented. The habit plane for hydride precipitation in Zr and U-Zr alloys is presented. The modulus of elasticity and thermal expansion of Zircaloy-2 between room temperature and 1000 F are reported. The tensile behavior of Zircaloy-2 as a function of heat treatment is discussed. The effect of neutron irradiation on the room- temperature tensile properties of Zircaloy-3 is reported. Process Metallurgy. Various processes were examined for producjng Zircaloy powder. A comparison is presented of cathodic discharge and chemical etching and the use of replication techniques for metallographic samples. The ultrasonic bond testing of narrow ligaments is reported. The joining of Al tubes with limited internal constriction in the weld areas is presented. (W.L.H.)

• Research Organization: Knolls Atomic Power Lab., Schenectady, N.Y.
• DOE Contract Number: W-31-109-ENG-52
• NSA Number: NSA-13-011173
• OSTI ID: 4279336
• Report Number(s): KAPL-2000-5
• Country of Publication: United States
• Language: English

Related Subjects

ADSORPTION
ALUMINUM
ALUMINUM ALLOYS
ALUMINUM OXIDES
ANNEALING
BORON CARBIDES
CADMIUM ALLOYS
CATHODES
CERAMICS
CERIUM OXIDES
CHROMIUM
CHROMIUM ALLOYS
CONTROL ELEMENTS
CORROSION
CORROSION PROTECTION
CRACKS
DEGASSING
DISPERSIONS
ELASTICITY
ELECTRON MICROSCOPY
ENRICHMENT
ETCHING
EXPANSION
EXTRUSION
FABRICATION
FISSION PRODUCTS
FOILS
FUEL ELEMENTS
FUELS
HEAT TREATMENTS
HIGH TEMPERATURE
HYDRIDES
HYDROGEN
IMPURITIES
INCONEL ALLOYS
INDIUM ALLOYS
IRRADIATION
JOINTS
LATTICES
MATHEMATICS
MECHANICAL PROPERTIES
MECHANICAL STRUCTURES
METALLOGRAPHY
METALLURGY
METALLURGY AND CERAMICS
MICROSCOPY
NEUTRON BEAMS
NICKEL ALLOYS