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Title: PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING DECEMBER, 1958

Abstract

Thermal-conductivity measurements were completed from 100 to 600 deg C on UO/sub 2/ specimens. An investigation of the creep properties of annealed and of 15% coldworked Zircaloy-2 in the 290 to 400 deg C temperature range is being conducted. Research to develop a method of sink-float density measurements to identify factors affecting irradiation-induced volume changes in graphite was continued. An increase in the U content of Al-U alloys is desirable as a means of increasing the fuel loading of reactors utilizing these fuels. The development of the radiometric method for the analysis of Mg in cement was completed. A study of the solidification of U castings in cylindrical graphite molds is being conducted. Experimental work was continued to reduce the amount of additive oxide (La/sub 2/O/sub 3/, CaO, and Y/sub 2/O/sub 3/) required to achieve stabilization of uranium oxides. Additional H/sub 2/-adsorption isotherms were obtained for the Zr-25 wt.% U alloy. An irradiation surveillance program is being conducted to determine the effect of fast neutrons on the mechanical properties of type 347 stainless steel. The study concerned with the properties of Nb-U alloys was continued. A program is being conducted to investigate methods for improving the irradiation behavior and corrosionmore » properties of Th -U alloys. An investigation is being made of cermet fuel materials consisting of from 60 to 90 vol. % of U0/sub 2/, UN, or UC dispersed in stainless steel. Gas-pressure bonding of Mo- and Nb-clad fuel elements is reported. A study to evaluate the irradiation resistance of dispersion fuel elements consisting of UC or UN fuel dispersed in a stainless steel matrix clad with stainless steel is reported. Fatigue studies of Inconel and INOR-8 are reported. A study of the constitution of U-Nb alloys is being conducted. Fundamental studies are being made of the kinetics and mechanism of the reaction of N/sub 2/ with Nb. The examination and evaluation of six 1.5-inch-diameter fueled (U0/sub 2/) graphite spheres after irradiation is reported. Localized attack on Ti steam tubes exposed to Darex dissolver solutions appears to be connected with defective areas in the tubes. Heat treatment of Ni-o-nel for 0.5 hr at 1850 deg F with an air quench following welding has produced the most resistance to Thorex solutions of any treatmert studied. The conditions prevailing during dissolution of Type 304 stainless steel in boiling 6 M H/sub 2/ SO/sub 4/ give corrosion rates in the range of 2 to 3 mils per month for Ni-o-nel specimens exposed to the vapor. Scouting experiments have shown the Zirflex decladding solution to be excessively corrosive to Type 347 and Carpenter 20 Cb stainless steels and Ni-o-nel. Raising the temperature from 650 to 700 deg C practically doubles the corrosiveness of the equimolar NaF-ZrF/sub 4/ salt. The evaluation of UC as a fuel for the SRE is proceeding. The amount of fission product Xe/sup 133/ released from UC as a function of time and temperature of post-irradiation heating was determined. The results of measurements of fission- gas release from Th-11 wt. % U specimers are reported. The effects of radiation and corrosion by the organic moderator on certain structural materials planned for use in the Organic Moderated Reactor are being studied. Properties of arc- melted Ta and Ta-W alloys are presented. A flat-plate Zircaloy-2-clad fuel element containing compartmented U0/sub 2 fuel is being considered for Core 2 of the PWR. The postirradiation examination of two stainless steel-clad U0/sub 2/ specimens for the Maritime Gas-Cooled Reactor program is continuing. Experiments leading to the development of a BeO--U0/sub 2/ fuel element were begun. (For preceding period see BMI-1304.) (W.L.H.)« less

Authors:
;
Publication Date:
Research Org.:
Battelle Memorial Inst., Columbus, Ohio
OSTI Identifier:
4234391
Report Number(s):
BMI-1307
NSA Number:
NSA-13-016196
DOE Contract Number:  
W-7405-ENG-92
Resource Type:
Technical Report
Resource Relation:
Other Information: Decl. June 12, 1959. Orig. Receipt Date: 31-DEC-59
Country of Publication:
United States
Language:
English
Subject:
METALLURGY AND CERAMICS; ADSORPTION; AIR; ALUMINUM ALLOYS; ANNEALING; BERYLLIUM OXIDES; BOILING; BONDING; BUILDING MATERIALS; CANNING; CASTING; CEMENTS; CERMETS; CHEMICAL REACTIONS; CHROMIUM ALLOYS; COLD WORKING; COPPER ALLOYS; CORROSION; CREEP; CYLINDERS; DAREX PROCESS; DENSITY; DOSEMETERS; ELECTRIC ARCS; FAST NEUTRONS; FATIGUE; FISSION PRODUCTS; FUEL ELEMENTS; FUELS; GAS COOLANT; GASES; GRAPHITE; HEAT TREATMENTS; HEATING; HIGH TEMPERATURE; HYDROGEN; INCONEL ALLOYS; INOR-8; IRON ALLOYS; IRRADIATION; MAGNESIUM; MEASURED VALUES; MECHANICAL PROPERTIES; MELTING; MIXING; MOLYBDENUM; MOLYBDENUM ALLOYS; NI-O-NEL; NICKEL ALLOYS; NIOBIUM; NIOBIUM ALLOYS; NITROGEN; ORGANIC MO

Citation Formats

Dayton, R.W., and Tipton, C.R. Jr. PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING DECEMBER, 1958. United States: N. p., 1959. Web. doi:10.2172/4234391.
Dayton, R.W., & Tipton, C.R. Jr. PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING DECEMBER, 1958. United States. doi:10.2172/4234391.
Dayton, R.W., and Tipton, C.R. Jr. Thu . "PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING DECEMBER, 1958". United States. doi:10.2172/4234391. https://www.osti.gov/servlets/purl/4234391.
@article{osti_4234391,
title = {PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING DECEMBER, 1958},
author = {Dayton, R.W. and Tipton, C.R. Jr.},
abstractNote = {Thermal-conductivity measurements were completed from 100 to 600 deg C on UO/sub 2/ specimens. An investigation of the creep properties of annealed and of 15% coldworked Zircaloy-2 in the 290 to 400 deg C temperature range is being conducted. Research to develop a method of sink-float density measurements to identify factors affecting irradiation-induced volume changes in graphite was continued. An increase in the U content of Al-U alloys is desirable as a means of increasing the fuel loading of reactors utilizing these fuels. The development of the radiometric method for the analysis of Mg in cement was completed. A study of the solidification of U castings in cylindrical graphite molds is being conducted. Experimental work was continued to reduce the amount of additive oxide (La/sub 2/O/sub 3/, CaO, and Y/sub 2/O/sub 3/) required to achieve stabilization of uranium oxides. Additional H/sub 2/-adsorption isotherms were obtained for the Zr-25 wt.% U alloy. An irradiation surveillance program is being conducted to determine the effect of fast neutrons on the mechanical properties of type 347 stainless steel. The study concerned with the properties of Nb-U alloys was continued. A program is being conducted to investigate methods for improving the irradiation behavior and corrosion properties of Th -U alloys. An investigation is being made of cermet fuel materials consisting of from 60 to 90 vol. % of U0/sub 2/, UN, or UC dispersed in stainless steel. Gas-pressure bonding of Mo- and Nb-clad fuel elements is reported. A study to evaluate the irradiation resistance of dispersion fuel elements consisting of UC or UN fuel dispersed in a stainless steel matrix clad with stainless steel is reported. Fatigue studies of Inconel and INOR-8 are reported. A study of the constitution of U-Nb alloys is being conducted. Fundamental studies are being made of the kinetics and mechanism of the reaction of N/sub 2/ with Nb. The examination and evaluation of six 1.5-inch-diameter fueled (U0/sub 2/) graphite spheres after irradiation is reported. Localized attack on Ti steam tubes exposed to Darex dissolver solutions appears to be connected with defective areas in the tubes. Heat treatment of Ni-o-nel for 0.5 hr at 1850 deg F with an air quench following welding has produced the most resistance to Thorex solutions of any treatmert studied. The conditions prevailing during dissolution of Type 304 stainless steel in boiling 6 M H/sub 2/ SO/sub 4/ give corrosion rates in the range of 2 to 3 mils per month for Ni-o-nel specimens exposed to the vapor. Scouting experiments have shown the Zirflex decladding solution to be excessively corrosive to Type 347 and Carpenter 20 Cb stainless steels and Ni-o-nel. Raising the temperature from 650 to 700 deg C practically doubles the corrosiveness of the equimolar NaF-ZrF/sub 4/ salt. The evaluation of UC as a fuel for the SRE is proceeding. The amount of fission product Xe/sup 133/ released from UC as a function of time and temperature of post-irradiation heating was determined. The results of measurements of fission- gas release from Th-11 wt. % U specimers are reported. The effects of radiation and corrosion by the organic moderator on certain structural materials planned for use in the Organic Moderated Reactor are being studied. Properties of arc- melted Ta and Ta-W alloys are presented. A flat-plate Zircaloy-2-clad fuel element containing compartmented U0/sub 2 fuel is being considered for Core 2 of the PWR. The postirradiation examination of two stainless steel-clad U0/sub 2/ specimens for the Maritime Gas-Cooled Reactor program is continuing. Experiments leading to the development of a BeO--U0/sub 2/ fuel element were begun. (For preceding period see BMI-1304.) (W.L.H.)},
doi = {10.2172/4234391},
journal = {},
number = ,
volume = ,
place = {United States},
year = {1959},
month = {1}
}