7 research outputs found
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Experience with stress corrosion cracking and materials compatibility at the High Flux Beam Reactor
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Correlation of high cycle and low cycle fatigue data for some HTGR structural metals
An analytical procedure has been evaluated to determine whether low and high cycle fatigue testing techniques may be correlated in the 10/sup 5/ cycle region where the data overlap. The procedure, which is based on the use of cyclic stress-strain curves to convert high cycle fatigue stresses to equivalent strains, is shown to be acceptable for Incoloy 800H, Hastelloy X, Type 304 stainless steel and 2 1/4 Cr--1Mo steel in the range of temperature for which data are available
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Effect of fission product interactions on the corrosion and mechanical properties of HTGR alloys. [HTGR]
Preliminary experiments have been carried out to determine how fission product interactions may influence the mechanical integrity of reference HTGR structural metals. In this work Type 304 stainless steel, Incoloy 800 and Hastelloy X were heated to 550 to 650/sup 0/C in the presence of CsI. It was found that no corrosion of the alloys occurred unless air or oxygen was also present. A mechanism for the observed behavior is proposed. A description is also given of some long term exposures of HTGR materials to more prototypic, low concentrations of I/sub 2/, Te/sub 2/ and CsI in the presence of low partial pressures of O/sub 2/. These samples are scheduled for mechanical bend tests after exposure to determine the degree of embrittlement
URANIUM-BISMUTH IN-PILE CORROSION TEST LOOP. RADIATION LOOP NO. 1
A loop was operated in the Brookhaven Graphite Research Reactor to determine the effect of in-pile irradiation on the corrosion of various materials by a U-- Bi solution. The loop wws fabricated of 21/4% chrome-1% Mo steel and contained, in the in-pile section, specimens of low-chrome steels, C steel, Mo, Be, Ta, and graphite. The U--Bi solution containing 869 ppm U/sup 235/ 98 ppm U/ sup 238/, 236 ppm Zr, and 346 ppm Mg was circulated at 51/4 gpm. A temperature difference of 75 deg C was maintained on the loop. The in-pile test section ran at 500 deg C and the finned cooler section at 425 deg C. The in-pile test section was exposed to a neutron flux of 4.4 x 10/sup 12/ neutrons/cm/sup 2/-sec which provided a fission density of 5.5 x 10/sup 10/ fissions/cm/sup 3/-sec. Metallographic examination indicated that the corrosion and/or erosion of the steel and graphite specimens was nil. Wetting of the specimens by the U-Bi solution was limited. Results indicate that in-pile and out-of-pile experimental results are similar and that fission fragment recoils did not contribute materially to either wetting or corrosion under the conditions imposed in this test. (auth