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Title: PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING JANUARY 1959

Abstract

Thermal-conductivity measurements are in progress on an unirradiated, unclad, natural uranium specimen. Thermal-conductivity measurements were made on medium-density UO/sub 2/. The creep properties of annealed and of 15% cold- worked Zircaloy-2 are being studied. Research to develop a method of sink-float density measurements to identify factors affecting radiationinduced volume changes in graphite was continued. A program was initiated to evaluate possible loss-ofcoolant incidents in the Plutonium Recycle Test Reactor (PRTR) by means of simulation on a digital computer. Research on the casting of hollow Al -35 wt.% U extrusion billets has been concerned with the utilization of bottom-pouring techniques in conjunction with a movable pouring spout feeding into a horizontal centrifugal mold. The development of a method for the analysis of Mg in cement is in progress. Infrared and gas-chromatography analyses of irradiated dodecane, decane, cetane, octane, and their urea complexes were continued. The manner in which U metal solidifies in cylindrical graphite molds is under study. Valence effects of oxide additions to UO/sub 2/ are being investigated. Progress in the study of potential fueled moderators has continued with the determination of hydrogenabsorption isotherms for the Zr-25 wt.% U alloy and high-temperature x- ray diffraction patterns of hydrides of thatmore » alloy. The effects of fast neutros flux on the mechanical properties of AISI Type 347 stainless steel are being determined and evaluated. Several Nbbase alloy systems are being considered for physicaland mechanical-property evaluation with respect to cladding applications in fast reactors. The fabrication of U-Nb alloys by various means is being studied. Thorium--uranium alloys are being studied for the purpose of developing improved corrosion resistance and irradiation stability of the alloys by means of alloying and control of processing variables. Cermet fuel materials consisting of from 60 to 90 vol.% UO/sub 2/, UN, or UC dispersed in a stainless steel or Nb matrix are being investigated. The gaa-pressure bonding technique is being investigated as a process for cladding fuel elements and assemblies with Mo and Nb. The irradiation behavior of UC-and UN-stainless steel dispersiontype fuel elements is being investigated. An experimental study of hydrogen migration in zirconium hydride under the influence of a thermal gradient is reported. The testing phases of the postirradiation examination of three pairs of fueled graphite spheres irradiated in capsule SP-1 were completed and results are reported. Titanium steam tubes are being exposed to initial and beginning Darex solutions. Data are presented on corrosion of INOR-1 and INOR-8 at various temperatures in a NaF-LiF-ZrFi salt. The postirradiation examination of twelve specimens of Th-11 wt. % U alloy was completed. Data are presented on the room- temperature tensile properties of arc-melted unalloyed Ta and nominal Ta-1.5 to 6.0 wt. % W alloys after vacuum annealing. The pressure bonding of Zircaloy-2- clad fuel elements containing compartmented oxide fuel plates is being investigated. (For preceding period see BMI-1307.) (W.L.H.)« less

Authors:
;
Publication Date:
Research Org.:
Battelle Memorial Inst., Columbus, Ohio
OSTI Identifier:
4228230
Report Number(s):
BMI-1315
NSA Number:
NSA-13-016970
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; ALKANES; ALUMINUM ALLOYS; ANNEALING; BONDING; CANNING; CASTING; CEMENTS; CENTRIFUGATION; CERMETS; CHROMATOGRAPHY; CHROMIUM ALLOYS; COLD WORKING; COMPUTERS; CONTROL; COOLANTS; CORROSION; CREEP; CYLINDERS; DAREX PROCESS; DENSITY; DIFFRACTION; DIFFUSION; DODECANE; ELECTRIC ARCS; EXTRUSION; FABRICATION; FAST NEUTRONS; FUEL CANS; FUEL ELEMENTS; GASES; GRAPHITE; HIGH TEMPERATURE; HYDRIDES; INFRARED RADIATION; INOR-8; LEAKS; LITHIUM FLUORIDES; MAGNESIUM; MATERIALS TESTING; MEASURED VALUES; MECHANICAL PROPERTIES; MOLYBDENUM; MOLYBDENUM ALLOYS; NICKEL ALLOYS; NIOBIUM; NIOBIUM ALLOYS; OCTANE; OXIDATION; PHASE DIAGRAMS; PLATES; PLUTONIUM; PR

Citation Formats

Dayton, R W, and Tipton, Jr, C R. PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING JANUARY 1959. United States: N. p., 1959. Web.
Dayton, R W, & Tipton, Jr, C R. PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING JANUARY 1959. United States.
Dayton, R W, and Tipton, Jr, C R. Sun . "PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING JANUARY 1959". United States.
@article{osti_4228230,
title = {PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING JANUARY 1959},
author = {Dayton, R W and Tipton, Jr, C R},
abstractNote = {Thermal-conductivity measurements are in progress on an unirradiated, unclad, natural uranium specimen. Thermal-conductivity measurements were made on medium-density UO/sub 2/. The creep properties of annealed and of 15% cold- worked Zircaloy-2 are being studied. Research to develop a method of sink-float density measurements to identify factors affecting radiationinduced volume changes in graphite was continued. A program was initiated to evaluate possible loss-ofcoolant incidents in the Plutonium Recycle Test Reactor (PRTR) by means of simulation on a digital computer. Research on the casting of hollow Al -35 wt.% U extrusion billets has been concerned with the utilization of bottom-pouring techniques in conjunction with a movable pouring spout feeding into a horizontal centrifugal mold. The development of a method for the analysis of Mg in cement is in progress. Infrared and gas-chromatography analyses of irradiated dodecane, decane, cetane, octane, and their urea complexes were continued. The manner in which U metal solidifies in cylindrical graphite molds is under study. Valence effects of oxide additions to UO/sub 2/ are being investigated. Progress in the study of potential fueled moderators has continued with the determination of hydrogenabsorption isotherms for the Zr-25 wt.% U alloy and high-temperature x- ray diffraction patterns of hydrides of that alloy. The effects of fast neutros flux on the mechanical properties of AISI Type 347 stainless steel are being determined and evaluated. Several Nbbase alloy systems are being considered for physicaland mechanical-property evaluation with respect to cladding applications in fast reactors. The fabrication of U-Nb alloys by various means is being studied. Thorium--uranium alloys are being studied for the purpose of developing improved corrosion resistance and irradiation stability of the alloys by means of alloying and control of processing variables. Cermet fuel materials consisting of from 60 to 90 vol.% UO/sub 2/, UN, or UC dispersed in a stainless steel or Nb matrix are being investigated. The gaa-pressure bonding technique is being investigated as a process for cladding fuel elements and assemblies with Mo and Nb. The irradiation behavior of UC-and UN-stainless steel dispersiontype fuel elements is being investigated. An experimental study of hydrogen migration in zirconium hydride under the influence of a thermal gradient is reported. The testing phases of the postirradiation examination of three pairs of fueled graphite spheres irradiated in capsule SP-1 were completed and results are reported. Titanium steam tubes are being exposed to initial and beginning Darex solutions. Data are presented on corrosion of INOR-1 and INOR-8 at various temperatures in a NaF-LiF-ZrFi salt. The postirradiation examination of twelve specimens of Th-11 wt. % U alloy was completed. Data are presented on the room- temperature tensile properties of arc-melted unalloyed Ta and nominal Ta-1.5 to 6.0 wt. % W alloys after vacuum annealing. The pressure bonding of Zircaloy-2- clad fuel elements containing compartmented oxide fuel plates is being investigated. (For preceding period see BMI-1307.) (W.L.H.)},
doi = {},
url = {https://www.osti.gov/biblio/4228230}, journal = {},
number = ,
volume = ,
place = {United States},
year = {1959},
month = {2}
}

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