PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING SEPTEMBER 1959
Tentative creep data are reported for annealed Zircaloy-2 sheet. A program directed toward the development of corrosion-resistant welding alloys for use with vacuum-melted low-carbon Hastelloy F to contain HAPO spent-fuel-element decladding solutions was initiatBe. Research to develop more satisfactory fuels from the Al--U system is reported. The development of a radiometric method for the determination of CaO in Port land cement was completed. In the study of radiation-induced nitration of hydrocarbons, a series of thermal and irradiation runs was completed in the liquid phase of the HNO/sub 3/-- cyclohexane determine the effects of ultra-high preasure and high temperature on uranium oxides and on the reactions of uranium oxides with mixed oxides. The irradiationBurveillance program was continued on type 347 stainless steel. Tensile data are reported for Nb-base alloys. A summary is reported of corrosion results obtained on Nb alloys exposed in high-temperature water and steam. The creep properties of Zircaloy-2 during irradiation at elevated temperatures are being investigated. Corrosion data are reported for Nb--U alloys exposed to high-temperature water and NaK. A program devoted to the determination of causes of fission-gas release in UO/sub 2/ is reported. Cermet and ceramic-type fuels are being clad with Mo and Nb by the gas-pressure-bonding technique. Data are reported on the densification of UO/sub 2/ by various pressure -bending conditions. Methods of producing dense UC pellets by powdermetallurgy methods are being investigated. Techniques for the production of high-quality cast shapes of UC are being developed. The rates of interdiffusion of U and C in the U-monocarbide-dicarbide system and the rates of selfdiffusion of U and C- in UC are being investigated. Hydrogen migration in Kr under the influence of a thermal gradient is being studied. Neutron-activation and in-pile experiments are being conducted to determine fission-gas retention and the effect on radiation on fueled-graphite spheres. Chemical analysis of cold-rolled binary Ta alloys is reported. Data are presented on the fission-gas release from UC -- graphite, UC/sub 2/-- graphite, and UO/sub 2/-- BeO specimens during postirradiation heat treatment in vacuum at 1800 deg F for 24 hours. Techniques are being developed for the fabrication of fuel elements, suppressor components, and control rods for the SM-2 reactor. Studies directed toward the development of compact gas--cooled reactors are reported. Research on core materials in support of the MGCR program is in progress. (For preceding period see BMI-1377.) (W.L.H.)
- Research Organization:
- Battelle Memorial Inst., Columbus, OH (United States)
- DOE Contract Number:
- W-7405-ENG-92
- NSA Number:
- NSA-14-014001
- OSTI ID:
- 4171617
- Report Number(s):
- BMI-1381
- Resource Relation:
- Other Information: Decl. Nov. 4, 1959. Orig. Receipt Date: 31-DEC-60
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
ACTIVATION- FISSION PRODUCTS- FUELS- GASES- GRAPHITE- NEUTRONS- QUANTITATIVE ANALYSIS- RADIATION EFFECTS- RETENTION- SPHERES
ALLOYS- ALUMINUM ALLOYS- CHROMIUM ALLOYS- COBALT ALLOYS- CORROSION- FUELS- HAPO- HASTELLOY- MELTING- MOLYBDENUM ALLOYS- NICKEL ALLOYS- STABILITY- URANIUM ALLOYS- USES- VACUUM- VESSELS- WASTE SOLUTIONS- WELDING
ANNEALING- CREEP- PLATES- ZIRCALOY
BERYLLIUM OXIDES- CARBIDES- FISSION PRODUCTS- GASES- - GRAPHITE- HEAT TREATMENTS- IRRADIATION- URANIUM COMPOUNDS- URANIUM DIOXIDE- VACUUM- WASTE DISPOSAL
BONDING- CANNING- CERAMICS- CERMETS- FUELS- GASES- MOLYBDENUM- NIOBIUM- PRES