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Title: PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING JULY 1959

Abstract

A study of the creep properties of 15% cold-worked and annealed Zircaloy- 2 sheet at 290, 345, and 400 deg C is continuing. Data are presented on the corrosion in bightemperature water of defected natural-U fuel cores extrusion clad with Zircaloy-2. The composition and temperature of the Al--U--Ni ternary eutectic in high-Al alloys are being determined. The study of neutron-activation analysis of cement raw materials was continued. Work continued on the study of radiation-induced changes in polymers leading to graft copolymerization. Work continued on the determination of valence effects of La/sub 2/O/sub 3/, Y/sub 2/O/ sub 3/, and CsO additi ons to UO/sub 2/. An investigation is being made of the effects of combined high pressure and high tsmperature on reactions of uranium oxide with various mixed oxides. A program is in progress to determine the effect of irradiation on the of uranium oxide with various mixed oxides. A program is in progress to determine the effect of irradiation on the mechanical properties of Al8l Type 347 stainless steel. A summary is presented of corrosion results obtained on Nb alloys exposed in high temperature water and steam. An investigation of the creep properties of Zircaloy-2 during irradiation at elevated temperaturesmore » continued. The properties of Nb-rich Nb--U alloys are being investigated. A program to develop Th-- U-bsee alloys of improved radiation sthbility by termary and quaternary alloying is in progress. Cement fuel materials consisting of from 60 to 90 vol.% UO/sub 2/ dispersed in Cr, Mo, Nb, or stainless steel matrices are being investigated. Work continued on the gaspressure bonding of Mo- and Nb-clad fuel elements. Coldcomparting studies were made with several grades of commercial UO/sub 2/ powder and results are reported. Fabrication of UC by various methods are reported. The rates of diffusion of U and C in UC are being investigated. The migration of hydrogen in Zr under the influence of a thermul gradient is being examined. Spherical fueled- graphite elements are being evaluated in preirradiation tests. Dath are reported on the fabricability and hardness of binary Ta alloys. Techniques were developed to coat by vapor deposition of Cr or pyrolynic carbon both blasted and unblasted UO/sub 2/ cores. Investigations have continued on the fabrication of Zircaloy-2- clad compartmented UO/sub 2/ fuel elements by gas-pressure bending. Rensely-sintered matrices of BeO containing about 20 vol.% of uniformly dispersed UO/sub 2/ particles are being investigated for fuel-element-core applica- . tions. Techniques for cladding UO/sub 2/ particles with impermeable shells of BeO are being developed. Data are reported on fission-gas release from UC/sub 2/-graphite and hours. An investigation is being conducted to develop fuel, absorbed, and suppressor materials for the SM-2. Progress of capsule-irradiation prograrus is reported for stsinless steel - UO/sub 2/ and-- UN dispersion fuel elements, solid UO/sub 2/ and annularly loaded UO/sub 2/ fuel pins, and graphits UO/sub 2/ are preub 3/, 20 vol.% evaluated for use in preparing fuel plates and absorber plates for the SM-2 reactor. Pellets of BeO--25 vol.% UO/sub 2/ are being prepared for loop and capsule exposures. (For preceding period see BMI- 1403.) (W.L.H.)« less

Authors:
;
Publication Date:
Research Org.:
Battelle Memorial Inst., Columbus, Ohio
OSTI Identifier:
4158238
Report Number(s):
BMI-1366
NSA Number:
NSA-14-018105
DOE Contract Number:  
W-7405-ENG-92
Resource Type:
Technical Report
Resource Relation:
Other Information: Decl. Feb. 4, 1960. Orig. Receipt Date: 31-DEC-60
Country of Publication:
United States
Language:
English
Subject:
METALS, CERAMICS, AND OTHER MATERIALS; ABSORPTION- FUELS- SM-2; ACTIVATION- CEMENTS- NEUTRONS; ADDITIVES- CALCIUM OXIDES- CHEMICAL REACTIONS- ELECTRONS- HIGH TEMPERATURE- IMPURITIES- LANTHANUM OXIDES- PRESSURE- URANIUM DIOXIDE- VALENCE- VALENCE ELECTRONS- YTTRIUM OXIDES; ALUMINUM ALLOYS- EUTECTICS- NICKEL ALLOYS- QUALITATIVE ANALYSIS- QUANTITATIVE ANALYSIS- TEMPERATURE- URANIUM ALLOYS; ANNEALING- COLD WORKING- CORROSION- CREEP- DEFECTS- - EXTRUSION- FUEL CANS- HIGH TEMPERATURE- NATURAL URANIUM FUEL- PLATES- WATER- ZIRCALOY; BERYLLIUM OXIDES- CARBIDES- FISSION- GASES- GRAPHITE- HEATING- URANIUM COMPOUNDS- URANIUM DIOXIDE- VACUUM- WASTE DISPOS

Citation Formats

Dayton, R.W., and Tipton, C.R. Jr. PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING JULY 1959. United States: N. p., 1959. Web. doi:10.2172/4158238.
Dayton, R.W., & Tipton, C.R. Jr. PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING JULY 1959. United States. doi:10.2172/4158238.
Dayton, R.W., and Tipton, C.R. Jr. Sat . "PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING JULY 1959". United States. doi:10.2172/4158238. https://www.osti.gov/servlets/purl/4158238.
@article{osti_4158238,
title = {PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING JULY 1959},
author = {Dayton, R.W. and Tipton, C.R. Jr.},
abstractNote = {A study of the creep properties of 15% cold-worked and annealed Zircaloy- 2 sheet at 290, 345, and 400 deg C is continuing. Data are presented on the corrosion in bightemperature water of defected natural-U fuel cores extrusion clad with Zircaloy-2. The composition and temperature of the Al--U--Ni ternary eutectic in high-Al alloys are being determined. The study of neutron-activation analysis of cement raw materials was continued. Work continued on the study of radiation-induced changes in polymers leading to graft copolymerization. Work continued on the determination of valence effects of La/sub 2/O/sub 3/, Y/sub 2/O/ sub 3/, and CsO additi ons to UO/sub 2/. An investigation is being made of the effects of combined high pressure and high tsmperature on reactions of uranium oxide with various mixed oxides. A program is in progress to determine the effect of irradiation on the of uranium oxide with various mixed oxides. A program is in progress to determine the effect of irradiation on the mechanical properties of Al8l Type 347 stainless steel. A summary is presented of corrosion results obtained on Nb alloys exposed in high temperature water and steam. An investigation of the creep properties of Zircaloy-2 during irradiation at elevated temperatures continued. The properties of Nb-rich Nb--U alloys are being investigated. A program to develop Th-- U-bsee alloys of improved radiation sthbility by termary and quaternary alloying is in progress. Cement fuel materials consisting of from 60 to 90 vol.% UO/sub 2/ dispersed in Cr, Mo, Nb, or stainless steel matrices are being investigated. Work continued on the gaspressure bonding of Mo- and Nb-clad fuel elements. Coldcomparting studies were made with several grades of commercial UO/sub 2/ powder and results are reported. Fabrication of UC by various methods are reported. The rates of diffusion of U and C in UC are being investigated. The migration of hydrogen in Zr under the influence of a thermul gradient is being examined. Spherical fueled- graphite elements are being evaluated in preirradiation tests. Dath are reported on the fabricability and hardness of binary Ta alloys. Techniques were developed to coat by vapor deposition of Cr or pyrolynic carbon both blasted and unblasted UO/sub 2/ cores. Investigations have continued on the fabrication of Zircaloy-2- clad compartmented UO/sub 2/ fuel elements by gas-pressure bending. Rensely-sintered matrices of BeO containing about 20 vol.% of uniformly dispersed UO/sub 2/ particles are being investigated for fuel-element-core applica- . tions. Techniques for cladding UO/sub 2/ particles with impermeable shells of BeO are being developed. Data are reported on fission-gas release from UC/sub 2/-graphite and hours. An investigation is being conducted to develop fuel, absorbed, and suppressor materials for the SM-2. Progress of capsule-irradiation prograrus is reported for stsinless steel - UO/sub 2/ and-- UN dispersion fuel elements, solid UO/sub 2/ and annularly loaded UO/sub 2/ fuel pins, and graphits UO/sub 2/ are preub 3/, 20 vol.% evaluated for use in preparing fuel plates and absorber plates for the SM-2 reactor. Pellets of BeO--25 vol.% UO/sub 2/ are being prepared for loop and capsule exposures. (For preceding period see BMI- 1403.) (W.L.H.)},
doi = {10.2172/4158238},
journal = {},
number = ,
volume = ,
place = {United States},
year = {1959},
month = {8}
}