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Title: PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING OCTOBER 1959

Abstract

The creep properties of 15% cold-worked Zircaloy-2, as evident from results being obtained, are superior to those of the annealed Zircaloy-2 at test temperatures of 290, 345, and 400 deg C. The development of a fuel-element leak detector was continued with studies to determine the effect of AgBr concentration and the flow rate of the I/sup 131/ solution -on the exchange between solid AgBr and I. Modification and improvement of a thermal-neutron-flux monitoring system for application to the Hanford reactors were continued. A program is reported for the development of corrosionresistant welding alloys for use with vacuum- melted lowcarbon Hastelloy F when used as a container material for spent fuel- element decladding solutions. Aluminumuranium alloys containing 0.5 to 3.0 wt.% Sn or Zr are being evaluated for possible use as reactor fuels. Additional activation analysis was performed on three samples of cement raw materials. Analyses for manganese and sodium were completed. Current work on the intrinsicradiotracer process-control application has dealt with the problem of Fe removal from refinery streams. An investigation is being conducted on the effects of oxide additions to UO/sub 2/, relative to slabillzation toward oxidation and volatilization. An irradiation-surveillance program concerned with the effects of irradiation onmore » the mechanical properties of AISI Type 347 stainless steel is reported. Evaluation is being made of several Nb-base alloys which should possess better properties than a V--10 wt.% Ti--1 wt.% Nb alloy (an acceptable cladding material for the EBR). The evaluation of selected Nb-base alloys for sevice in pressurizedwater reactors was continued. An investigation of the creep properties of Zircaloy-2 during irradiation at elevated temperatures is being made. An investigation of the fabrication characteristics, mechanical and physical properties, and corrosion behavior in various media of Nb-- U alloys is being made to determine their applicability as reactor fuels. Thorium - uranium and thorium--uranium base alloys are being investigated with the aim of improving their irradiation stability and corrosion resistance. A program aimed at investigating Pu-alloy systems for possible fuel materials is being planned. In an attempt to develop alloys for application at reasonably high temperatures, Zr--Nb--Pu and U--Nb--Pu alloys will be investigated. Preparations for study of the important mechanisms of gas release in UO/sub 2/ in progress. This study will include both a determination of diffusion coefficients in UO/sub 2/ specimens of known geometry and use of these data in an in-pile study of gas release from sintered UO/sub 2/. Evaluations of fabrication techniques for producing UO/sub 2/- Mo and UO/sub 2/--Nb cermets of 90% theoretical density or better containing 60 to 90 vol.% fuel are being made. Investigations of gas- pressure bonding of Mo- and Nb-clad fuel elements were continued. The uranium carbides and particularly uranium monocarbide are being developed for use as power reactor fuels. A program is reported for the preparation of UC bodies by powder metallurgy processes. Reliable techniques for the production of high- quality cast shapes of UC are being developed. A fundamental investigation is being made of the reactions of N/sub 2/ with Nb. A program was initiated to produce UO/sub 2/ crystals by the process of vaporization. Another investigation was made to grow monocrystals of UO/sub 2/ by the use of fusion methods. Research on core materials in support of the MGCR program is in progress. The major effort is on the development and evaluation of UO/sub 2/ dispersions in BeO and dispersions of UC and UC/sub 2/ in graphite, and on the cladding of UO/sub 2/ particles with BeO. The development of fuel, absorbers, and suppressor materials for the SM-2 is reported. (For preceding period see BMI-1381.) (W.L.H.)« less

Authors:
;
Publication Date:
Research Org.:
Battelle Memorial Inst., Columbus, Ohio
OSTI Identifier:
4160919
Report Number(s):
BMI-1391(Rev.)
NSA Number:
NSA-14-014002
DOE Contract Number:  
W-7405-ENG-92
Resource Type:
Technical Report
Resource Relation:
Other Information: 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; ADDITIVES- ALUMINUM ALLOYS- FUELS- TIN- URANIUM ALLOYS- USES- ZIRCONIUM; ADDITIVES- EFFICIENCY- OXIDATION- OXIDES- STABILITY- URANIUM DIOXIDE- VOLATILITY; ALLOYS- CHROMIUM ALLOYS- COBALT ALLOYS- CORROSION- HASTELLOY- MELTING- MOLYBDENUM ALLOYS- NICKEL ALLOYS- STABILITY- USES- VACUUM- VESSELS- WASTE SOLUTIONS- WELDING; ANNEALING- COLD WORKING- CREEP- TEMPERATURE- ZIRCALOY; BERYLLIUM OXIDES- CARBIDES- DISPERSIONS- - FUEL CANS- GRAPHITE- MGCR- URANIUM COMPOUNDS- URANIUM DIOXIDE; BONDING- FUEL CANS- GASES- MOLYBDENUM- NIOBIUM- PRESSURE; CARBIDES- FUELS- POW

Citation Formats

Dayton, R.W., and Tipton, C.R. Jr. PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING OCTOBER 1959. United States: N. p., 1959. Web. doi:10.2172/4160919.
Dayton, R.W., & Tipton, C.R. Jr. PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING OCTOBER 1959. United States. doi:10.2172/4160919.
Dayton, R.W., and Tipton, C.R. Jr. Sun . "PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING OCTOBER 1959". United States. doi:10.2172/4160919. https://www.osti.gov/servlets/purl/4160919.
@article{osti_4160919,
title = {PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING OCTOBER 1959},
author = {Dayton, R.W. and Tipton, C.R. Jr.},
abstractNote = {The creep properties of 15% cold-worked Zircaloy-2, as evident from results being obtained, are superior to those of the annealed Zircaloy-2 at test temperatures of 290, 345, and 400 deg C. The development of a fuel-element leak detector was continued with studies to determine the effect of AgBr concentration and the flow rate of the I/sup 131/ solution -on the exchange between solid AgBr and I. Modification and improvement of a thermal-neutron-flux monitoring system for application to the Hanford reactors were continued. A program is reported for the development of corrosionresistant welding alloys for use with vacuum- melted lowcarbon Hastelloy F when used as a container material for spent fuel- element decladding solutions. Aluminumuranium alloys containing 0.5 to 3.0 wt.% Sn or Zr are being evaluated for possible use as reactor fuels. Additional activation analysis was performed on three samples of cement raw materials. Analyses for manganese and sodium were completed. Current work on the intrinsicradiotracer process-control application has dealt with the problem of Fe removal from refinery streams. An investigation is being conducted on the effects of oxide additions to UO/sub 2/, relative to slabillzation toward oxidation and volatilization. An irradiation-surveillance program concerned with the effects of irradiation on the mechanical properties of AISI Type 347 stainless steel is reported. Evaluation is being made of several Nb-base alloys which should possess better properties than a V--10 wt.% Ti--1 wt.% Nb alloy (an acceptable cladding material for the EBR). The evaluation of selected Nb-base alloys for sevice in pressurizedwater reactors was continued. An investigation of the creep properties of Zircaloy-2 during irradiation at elevated temperatures is being made. An investigation of the fabrication characteristics, mechanical and physical properties, and corrosion behavior in various media of Nb-- U alloys is being made to determine their applicability as reactor fuels. Thorium - uranium and thorium--uranium base alloys are being investigated with the aim of improving their irradiation stability and corrosion resistance. A program aimed at investigating Pu-alloy systems for possible fuel materials is being planned. In an attempt to develop alloys for application at reasonably high temperatures, Zr--Nb--Pu and U--Nb--Pu alloys will be investigated. Preparations for study of the important mechanisms of gas release in UO/sub 2/ in progress. This study will include both a determination of diffusion coefficients in UO/sub 2/ specimens of known geometry and use of these data in an in-pile study of gas release from sintered UO/sub 2/. Evaluations of fabrication techniques for producing UO/sub 2/- Mo and UO/sub 2/--Nb cermets of 90% theoretical density or better containing 60 to 90 vol.% fuel are being made. Investigations of gas- pressure bonding of Mo- and Nb-clad fuel elements were continued. The uranium carbides and particularly uranium monocarbide are being developed for use as power reactor fuels. A program is reported for the preparation of UC bodies by powder metallurgy processes. Reliable techniques for the production of high- quality cast shapes of UC are being developed. A fundamental investigation is being made of the reactions of N/sub 2/ with Nb. A program was initiated to produce UO/sub 2/ crystals by the process of vaporization. Another investigation was made to grow monocrystals of UO/sub 2/ by the use of fusion methods. Research on core materials in support of the MGCR program is in progress. The major effort is on the development and evaluation of UO/sub 2/ dispersions in BeO and dispersions of UC and UC/sub 2/ in graphite, and on the cladding of UO/sub 2/ particles with BeO. The development of fuel, absorbers, and suppressor materials for the SM-2 is reported. (For preceding period see BMI-1381.) (W.L.H.)},
doi = {10.2172/4160919},
journal = {},
number = ,
volume = ,
place = {United States},
year = {1959},
month = {11}
}