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Title: PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING AUGUST 1959

Abstract

Data are presented on the creep properties of 15% coldworked Zircaloy-2 at 290, 345, and 400 deg C. The progress on the development of an isotopic- exchange fuel-element leak-detection system is summarized. A program to develop a thermal-neutron-flux monitoring system for Hanford reactors is reported. Work on the radiometric analysis of calcium in cement was continued. A surveillance program is in progress to determine the effects of fastneutron irradiation on the mechanical properties of AISI Type 347 stainless steel. A summary of corrosion results obtained on Nb, Nb-Zr, Nb-W, Nb-Mo, Nb-V, Nb-Fe, Nb-Ti, Nb-Ti-Cr, and Nb- Ti-V alloys exposed in hightemperature water and steam is presented. A study of the creep characteristics of Zircaloy-2 at elevated temperatures during exposure to a fast-neutron flux was initiated. A research program was initiated to develop an analytical technique for monitoring the oxygen concentration in largescale radioactive sodium systems. Data are presented on the corrosion behavior cf Nb-U alloys in high-temperature water after 98 days. The development of Th -U-base alloys of improved radiation stability and corrosion resistance is reported. Methods of producing cermets of 90% of theoretical density or better containing 60 to 90 vol.% ceramic fuel are being investigated. The gas-pressure- bondingmore » technique is being investigated as a possible method for fabricating Mo- and Nb-clad ceramic and cermet-type fuels. Data are presented on the densification and green compactibility of UO/sub 2/ powders. A summary of tensile-test results for pressure-bonded type 304 stainless steel bond speciments is presented. An investigation of methods for producing dense UC cores is being made. Data are presented on the effect of heat treatment at 1500 deg C for 1 hr on density and rupture strength of UC. The fission gas release from irradiated fueled-graphite spheres was studied. Thermal-conductivity measurements were completed on two UC specimens prepared by casting. Work on a series of binary Ta- base alloys containing additions of Hf, Th, Ti, W, Y, and Zr is nearing completion. Fuel-element cores consisting of approximately 20 vol.% UO/sub 2/ particles dispersed uniformly in a densely sintered BeO matrix are being developed. Research to develop improved graphite fuelelement cores containing UC or UC/sub 2/ particles in an amount equivalent to 20 vol.% of UO/sub 2/ is reported. Data are pre sented on the fission-gas release from 48.25 wt.% BeOUO/ sub 2/ pellets during postirradiation heat treatment in vacuum at 1800 and 2000 deg F for 24 hr. A study of the radiation stability of ceramic-type fuels under conditions simulating those of MGCR design is in progress. Research concerned with the development of fuel, absorber, and suppressor materials for the SM-2 is reported. In the GasCooled Reactor Program progress of capsule-irradiation is reported for stainless steel-UO/sub 2/ and -UN dispersion fuel elements, solid UO/ sub 2/ and annularly loaded UO/sub 2/ fuel pins, and graphite UO/sub 2/ fuel bodies. (For preceding period see BMI-1366.) (W.L.H.)« less

Authors:
;
Publication Date:
Research Org.:
Battelle Memorial Inst., Columbus, Ohio
OSTI Identifier:
4153427
Report Number(s):
BMI-1377
NSA Number:
NSA-14-016492
DOE Contract Number:  
W-7405-ENG-92
Resource Type:
Technical Report
Resource Relation:
Other Information: Decl. Oct. 8, 1959. Orig. Receipt Date: 31-DEC-60
Country of Publication:
United States
Language:
English
Subject:
GENERAL AND MISCELLANEOUS; BERYLLIUM OXIDES- CONFIGURATION- DENSITY- DISPERSIONS- FISSION PRODUCTS- FUEL ELEMENTS- - GASES- HEATING- HIGH TEMPERATURE- LOSSES- MEASURED VALUES- PELLETS- PLANNING- REACTOR CORE- REACTORS- SINTERED MATERIALS- URANIUM OXIDES- VACUUM; BONDING- CERAMICS- CERMETS- DENSITY- FABRICATION- FUELS- GASES- IRRADIATION- MGCR- MOLYBDENUM- NIOBIUM- PLANNING- PRESSURE- REACTORS; BONDING- MATERIALS TESTING- MEASURED VALUES- PRESSURE- REACTORS- STAINLESS STEELS- TENSILE PROPERTIES; CALCIUM- CEMENTS- RADIATION DETECTORS- REACTORS; CAPSULES- DISPERSIONS- FUEL ELEMENTS- GAS COOLANT- GRAPHITE- IRRADIATION- MATERIALS

Citation Formats

Dayton, R.W., and Tipton, C.R. Jr. PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING AUGUST 1959. United States: N. p., 1959. Web. doi:10.2172/4153427.
Dayton, R.W., & Tipton, C.R. Jr. PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING AUGUST 1959. United States. doi:10.2172/4153427.
Dayton, R.W., and Tipton, C.R. Jr. Tue . "PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING AUGUST 1959". United States. doi:10.2172/4153427. https://www.osti.gov/servlets/purl/4153427.
@article{osti_4153427,
title = {PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING AUGUST 1959},
author = {Dayton, R.W. and Tipton, C.R. Jr.},
abstractNote = {Data are presented on the creep properties of 15% coldworked Zircaloy-2 at 290, 345, and 400 deg C. The progress on the development of an isotopic- exchange fuel-element leak-detection system is summarized. A program to develop a thermal-neutron-flux monitoring system for Hanford reactors is reported. Work on the radiometric analysis of calcium in cement was continued. A surveillance program is in progress to determine the effects of fastneutron irradiation on the mechanical properties of AISI Type 347 stainless steel. A summary of corrosion results obtained on Nb, Nb-Zr, Nb-W, Nb-Mo, Nb-V, Nb-Fe, Nb-Ti, Nb-Ti-Cr, and Nb- Ti-V alloys exposed in hightemperature water and steam is presented. A study of the creep characteristics of Zircaloy-2 at elevated temperatures during exposure to a fast-neutron flux was initiated. A research program was initiated to develop an analytical technique for monitoring the oxygen concentration in largescale radioactive sodium systems. Data are presented on the corrosion behavior cf Nb-U alloys in high-temperature water after 98 days. The development of Th -U-base alloys of improved radiation stability and corrosion resistance is reported. Methods of producing cermets of 90% of theoretical density or better containing 60 to 90 vol.% ceramic fuel are being investigated. The gas-pressure- bonding technique is being investigated as a possible method for fabricating Mo- and Nb-clad ceramic and cermet-type fuels. Data are presented on the densification and green compactibility of UO/sub 2/ powders. A summary of tensile-test results for pressure-bonded type 304 stainless steel bond speciments is presented. An investigation of methods for producing dense UC cores is being made. Data are presented on the effect of heat treatment at 1500 deg C for 1 hr on density and rupture strength of UC. The fission gas release from irradiated fueled-graphite spheres was studied. Thermal-conductivity measurements were completed on two UC specimens prepared by casting. Work on a series of binary Ta- base alloys containing additions of Hf, Th, Ti, W, Y, and Zr is nearing completion. Fuel-element cores consisting of approximately 20 vol.% UO/sub 2/ particles dispersed uniformly in a densely sintered BeO matrix are being developed. Research to develop improved graphite fuelelement cores containing UC or UC/sub 2/ particles in an amount equivalent to 20 vol.% of UO/sub 2/ is reported. Data are pre sented on the fission-gas release from 48.25 wt.% BeOUO/ sub 2/ pellets during postirradiation heat treatment in vacuum at 1800 and 2000 deg F for 24 hr. A study of the radiation stability of ceramic-type fuels under conditions simulating those of MGCR design is in progress. Research concerned with the development of fuel, absorber, and suppressor materials for the SM-2 is reported. In the GasCooled Reactor Program progress of capsule-irradiation is reported for stainless steel-UO/sub 2/ and -UN dispersion fuel elements, solid UO/ sub 2/ and annularly loaded UO/sub 2/ fuel pins, and graphite UO/sub 2/ fuel bodies. (For preceding period see BMI-1366.) (W.L.H.)},
doi = {10.2172/4153427},
journal = {},
number = ,
volume = ,
place = {United States},
year = {1959},
month = {9}
}