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Title: PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING MAY 1959

Abstract

Data are presented on the electrical conductivity of unirradiated UO/sub 2/. A stud of the creep properties of annealed and of 15% cold-worked Zircaloy-2 at 290, 345, and 400 deg C is in progress. High-purity Mo single crystals are being grown in a modified Andrade furnace. Research to develop a method of sink- float density measurement to identify factors affecting irradiation-induced volume changes in graphite was continued. The simulation of conditions after a lossof-coolant incident in the PRTR by means of a digital computer was continued. Tensile data are given on Al--as wt. % U alloys with 3 wt. % ternary additions of Si, Sn, and Zr. Work was continued on the free-radical and peroxy grafting of polymethylmethacrylate and polyacetylmethacrylate. The manner in which uranium solidifies in cylindrical graphite molds is being studied. An investigation is being made of the effect of oxide additions (CaO and La/sub 2/O/sub 3/) to UO/sub 2/ in preventing the transformation and volatility associated with the formation of U/sub 3/O/sub 8/ during oxidation. Postirradiation studies are in progress on the specimens of hydrided Zr--2 wt. % U which were irradiated in the MTR. An irradiation-surveillance program is being conducted on AISI Type 347 stainless steel.more » An evaluation of several Nb-base alloys for possible application as alternate cladding material in future EBR cores is being made. Corrosion data are presented for commercial purity Nb-Zr, Nb-W, Nb-Mo, Nb-V, Nb-Fe, Nb-Ti, Nb-Ti- Cr-, Nb-Ti-Mo, and Nb-Ti-V alloys in 750 deg F steam and 650 deg F water. Corrosion data are reported for Nb-U alloys in air and in water after 14 and 25 days exposure. Thermal analysis data are presented for Th--U--Zr and Th--U--Zr-- Nb alloys. Specimens of single-crystal UO/sub 2/ are being prepared for study of fission-gas diffusion coefficients by low temperature irradiation and subsequent heating to collect the fission gases. The gas-pressure bonding technique is being studied as a possible method for cladding ceramic and cermet fuels with Mo and Nb. The high-temperature irradiation behavior of dispersion fuel elements consisting of 24 wt. % UN or UC dispersed in stainless steel and clad with stainless steel is being evaluated. The solid-state bonding of metals by application of heat and pressure is being studied in an attempt to establish the mechanism and kinetics of the process. Data are presented on the green compacting of U% powders under various conditions. Several methods of fabricating UC of high density are being studied. Data are presented on the effects of impurities on some properties of cast UC. A program to obtain basic fatigue information on Inconel is presented. A program concerned with the investigation of the temperature and frequency dependence of fatigue properties of INOR-8 alloy is reported Preliminary experiments were conducted with the apparatus designed for the study-of the migration of hydrogen in Zr under the influence of a thermal gradient. Fueled-graphite spheres are being evaluated with respect to selected thermal and mechanical properties. Irradiation data are presented on UC specimens. A Zircaloy-2-clad Compartmented flat-plate fuel element containg UO/sub 2/ cores is being considered for the PWR. Core-2. Techniques are being investigated for use in the fabrication of fuel elements containing cores of 26 wt. % UO/sub 2/-1.1 wt. % ZrB/sub 2/ dispersed in a prealloyed Type 347 stainless steel matrix. (For preceding period see BMI-1340.) (W.L.H.)« less

Authors:
;
Publication Date:
Research Org.:
Battelle Memorial Inst., Columbus, Ohio
OSTI Identifier:
4220558
Report Number(s):
BMI-1346
NSA Number:
NSA-13-021169
DOE Contract Number:  
W-7405-ENG-92
Resource Type:
Technical Report
Resource Relation:
Other Information: Decl. July 17, 1959. Orig. Receipt Date: 31-DEC-59
Country of Publication:
United States
Language:
English
Subject:
METALLURGY AND CERAMICS; ACCIDENTS- COMPUTERS- COOLANTS- DIGITAL SYSTEMS- LEAKS- PRTR- REACTORS; ACETYL RADICALS- FREE RADICALS- METHACRYLATES- METHACRYLIC ACID- METHYL RADICALS- POLYMERS; ALUMINUM ALLOYS- MEASURED VALUES- SILICIDES- TENSILE PROPERTIES- TIN ALLOYS- URANIUM ALLOYS- ZIRCONIUM ALLOYS; ANNEALING- COLD WORKING- CREEP- MEASURED VALUES- TEMPERATURE- TENSILE PROPERTIES- ZIRCALOY; BONDING- CANNING- CERAMICS- CERMETS- FUELS- GASES- MOLYBDENUM- NIOBIUM- PRESSURE; BONDING- HEATING- METALS- PRESSURE; CALCIUM OXIDES- CHEMICAL REACTIONS- LANTHANUM OXIDES- OXIDATION- URANIUM DIOXIDE- U3O8- VOLATILITY; CASTING- CYLINDERS- GRA

Citation Formats

Dayton, R.W., and Tipton, C.R. Jr. PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING MAY 1959. United States: N. p., 1959. Web. doi:10.2172/4220558.
Dayton, R.W., & Tipton, C.R. Jr. PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING MAY 1959. United States. doi:10.2172/4220558.
Dayton, R.W., and Tipton, C.R. Jr. Mon . "PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING MAY 1959". United States. doi:10.2172/4220558. https://www.osti.gov/servlets/purl/4220558.
@article{osti_4220558,
title = {PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING MAY 1959},
author = {Dayton, R.W. and Tipton, C.R. Jr.},
abstractNote = {Data are presented on the electrical conductivity of unirradiated UO/sub 2/. A stud of the creep properties of annealed and of 15% cold-worked Zircaloy-2 at 290, 345, and 400 deg C is in progress. High-purity Mo single crystals are being grown in a modified Andrade furnace. Research to develop a method of sink- float density measurement to identify factors affecting irradiation-induced volume changes in graphite was continued. The simulation of conditions after a lossof-coolant incident in the PRTR by means of a digital computer was continued. Tensile data are given on Al--as wt. % U alloys with 3 wt. % ternary additions of Si, Sn, and Zr. Work was continued on the free-radical and peroxy grafting of polymethylmethacrylate and polyacetylmethacrylate. The manner in which uranium solidifies in cylindrical graphite molds is being studied. An investigation is being made of the effect of oxide additions (CaO and La/sub 2/O/sub 3/) to UO/sub 2/ in preventing the transformation and volatility associated with the formation of U/sub 3/O/sub 8/ during oxidation. Postirradiation studies are in progress on the specimens of hydrided Zr--2 wt. % U which were irradiated in the MTR. An irradiation-surveillance program is being conducted on AISI Type 347 stainless steel. An evaluation of several Nb-base alloys for possible application as alternate cladding material in future EBR cores is being made. Corrosion data are presented for commercial purity Nb-Zr, Nb-W, Nb-Mo, Nb-V, Nb-Fe, Nb-Ti, Nb-Ti- Cr-, Nb-Ti-Mo, and Nb-Ti-V alloys in 750 deg F steam and 650 deg F water. Corrosion data are reported for Nb-U alloys in air and in water after 14 and 25 days exposure. Thermal analysis data are presented for Th--U--Zr and Th--U--Zr-- Nb alloys. Specimens of single-crystal UO/sub 2/ are being prepared for study of fission-gas diffusion coefficients by low temperature irradiation and subsequent heating to collect the fission gases. The gas-pressure bonding technique is being studied as a possible method for cladding ceramic and cermet fuels with Mo and Nb. The high-temperature irradiation behavior of dispersion fuel elements consisting of 24 wt. % UN or UC dispersed in stainless steel and clad with stainless steel is being evaluated. The solid-state bonding of metals by application of heat and pressure is being studied in an attempt to establish the mechanism and kinetics of the process. Data are presented on the green compacting of U% powders under various conditions. Several methods of fabricating UC of high density are being studied. Data are presented on the effects of impurities on some properties of cast UC. A program to obtain basic fatigue information on Inconel is presented. A program concerned with the investigation of the temperature and frequency dependence of fatigue properties of INOR-8 alloy is reported Preliminary experiments were conducted with the apparatus designed for the study-of the migration of hydrogen in Zr under the influence of a thermal gradient. Fueled-graphite spheres are being evaluated with respect to selected thermal and mechanical properties. Irradiation data are presented on UC specimens. A Zircaloy-2-clad Compartmented flat-plate fuel element containg UO/sub 2/ cores is being considered for the PWR. Core-2. Techniques are being investigated for use in the fabrication of fuel elements containing cores of 26 wt. % UO/sub 2/-1.1 wt. % ZrB/sub 2/ dispersed in a prealloyed Type 347 stainless steel matrix. (For preceding period see BMI-1340.) (W.L.H.)},
doi = {10.2172/4220558},
journal = {},
number = ,
volume = ,
place = {United States},
year = {1959},
month = {6}
}