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Title: PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING NOVEMBER 1958

Abstract

The effect of irradiation on the thermal conductivity and electrical resistivity of U and U0/sub 2/ is being investigated. The creep properties of 15% cold-worked Zircaloy-2 are being investigated in the 290 to 400 deg C temperature range for times exceeding 10,000 hr. The density distribution of crushed graphite is being investigated by the sink-float method. Centrifugal- casting techniques for the production of Al-35 wt.% U casting in the form of hollow cylinders are being investigated. A study of the processes involved in the solidification of U castings in graphite molds is being made. Work continued on electrolytic oxide and electroless oxide coatings on Croloy-2 1/4. Experimental work was continued to determine the effect of additive oxides on the oxidation characteristics and phase stability of U0/sub 2/. The fueled-moderator study has continued with the determination of additional hydrogen-absorption isotherms for the Zr-25 wt.% alloy and high-temperature x-ray diffraction patterns of hydrides of the 1 and 50 wt.% alloys. The irradiation of type 347 stainless steel at ETR process-water temperature. about 120 deg F and at 600 deg F, and subsequent determination of irradiation damage are being done in support of the KAPL-33 loop to be installed at the ETR. AIloysmore » of U and Nb are being considered as possible high-temperature reactor fuels. Thorium-uranium base alloys are the subject of an investigation aimed at improving irradiation stability and corrosion resistance by ternary alloying and control of processing techniques Cermet fuel materials consisting of from 60 to 90 vol.% U0/sub 2/, UN, or UC dispersed in stainless steel are being investigated. Several types of 1 1/ 2-inch-diameter fueled graphite spheres containing 10 wt.% of fully enriched U0/ sub 2/ are being evaluated before and after irradiation in the BRR. Some localized attack has been observed after prolonged exposure of Ti steam tubes to initial and beginning Darex solutions. Heat treating stabilized Nionel, prior to welding, has improved the resistance of weldments to initial Thorex solutions. Experiments are being conducted to determine the suitability of UC as a fuel for the SRE. Reports of progress are included in the irradiation, postirradiation- examination, and gas-releasing study of various capsules and specimens. All parts of the program for postirradiation examination of Th-11 wt.% U specimens have been completed except for metallography. Postirradiation inspection of the U0/sub 2/ fuel pins has been completed. Postirradiation examination of the 25 wt.% U0/sub 2/-stainless steel fuel element, OMRE-3. has been completed except for additional metallographic examinations. In the container-material research. high-quality Ta and Ta-W specimens have been prepared by arc melting, and strip- fabrication operations on the specimens have started. A compartmentalized Zircaloy-2-clad flat fuel element containing U0/sub 2/ cores is being considered for the PWR Core 2. A one-capsule irradiation program to evaluate the compatibility of graphite and CO/sub 2/ with types 310 and 446 stainless steels and with Inconel is nearing completion. (For preceding period see BMI-130l.) (W.L.H.)« less

Authors:
;
Publication Date:
Research Org.:
Battelle Memorial Inst., Columbus, Ohio
OSTI Identifier:
4252484
Report Number(s):
BMI-1304
NSA Number:
NSA-13-007729
DOE Contract Number:  
W-7405-ENG-92
Resource Type:
Technical Report
Resource Relation:
Other Information: Decl. Jan. 7, 1959. Orig. Receipt Date: 31-DEC-59
Country of Publication:
United States
Language:
English
Subject:
METALLURGY AND CERAMICS; DENSITY; DISTRIBUTION; FLOTATION; GRAPHITE

Citation Formats

Dayton, R.W., and Tipton, C.R. Jr. PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING NOVEMBER 1958. United States: N. p., 1958. Web. doi:10.2172/4252484.
Dayton, R.W., & Tipton, C.R. Jr. PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING NOVEMBER 1958. United States. doi:10.2172/4252484.
Dayton, R.W., and Tipton, C.R. Jr. Mon . "PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING NOVEMBER 1958". United States. doi:10.2172/4252484. https://www.osti.gov/servlets/purl/4252484.
@article{osti_4252484,
title = {PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING NOVEMBER 1958},
author = {Dayton, R.W. and Tipton, C.R. Jr.},
abstractNote = {The effect of irradiation on the thermal conductivity and electrical resistivity of U and U0/sub 2/ is being investigated. The creep properties of 15% cold-worked Zircaloy-2 are being investigated in the 290 to 400 deg C temperature range for times exceeding 10,000 hr. The density distribution of crushed graphite is being investigated by the sink-float method. Centrifugal- casting techniques for the production of Al-35 wt.% U casting in the form of hollow cylinders are being investigated. A study of the processes involved in the solidification of U castings in graphite molds is being made. Work continued on electrolytic oxide and electroless oxide coatings on Croloy-2 1/4. Experimental work was continued to determine the effect of additive oxides on the oxidation characteristics and phase stability of U0/sub 2/. The fueled-moderator study has continued with the determination of additional hydrogen-absorption isotherms for the Zr-25 wt.% alloy and high-temperature x-ray diffraction patterns of hydrides of the 1 and 50 wt.% alloys. The irradiation of type 347 stainless steel at ETR process-water temperature. about 120 deg F and at 600 deg F, and subsequent determination of irradiation damage are being done in support of the KAPL-33 loop to be installed at the ETR. AIloys of U and Nb are being considered as possible high-temperature reactor fuels. Thorium-uranium base alloys are the subject of an investigation aimed at improving irradiation stability and corrosion resistance by ternary alloying and control of processing techniques Cermet fuel materials consisting of from 60 to 90 vol.% U0/sub 2/, UN, or UC dispersed in stainless steel are being investigated. Several types of 1 1/ 2-inch-diameter fueled graphite spheres containing 10 wt.% of fully enriched U0/ sub 2/ are being evaluated before and after irradiation in the BRR. Some localized attack has been observed after prolonged exposure of Ti steam tubes to initial and beginning Darex solutions. Heat treating stabilized Nionel, prior to welding, has improved the resistance of weldments to initial Thorex solutions. Experiments are being conducted to determine the suitability of UC as a fuel for the SRE. Reports of progress are included in the irradiation, postirradiation- examination, and gas-releasing study of various capsules and specimens. All parts of the program for postirradiation examination of Th-11 wt.% U specimens have been completed except for metallography. Postirradiation inspection of the U0/sub 2/ fuel pins has been completed. Postirradiation examination of the 25 wt.% U0/sub 2/-stainless steel fuel element, OMRE-3. has been completed except for additional metallographic examinations. In the container-material research. high-quality Ta and Ta-W specimens have been prepared by arc melting, and strip- fabrication operations on the specimens have started. A compartmentalized Zircaloy-2-clad flat fuel element containing U0/sub 2/ cores is being considered for the PWR Core 2. A one-capsule irradiation program to evaluate the compatibility of graphite and CO/sub 2/ with types 310 and 446 stainless steels and with Inconel is nearing completion. (For preceding period see BMI-130l.) (W.L.H.)},
doi = {10.2172/4252484},
journal = {},
number = ,
volume = ,
place = {United States},
year = {1958},
month = {12}
}