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160219s1970 pau|||||o|||||||||||eng|| |
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|a 10.1520/STP26646S
|2 doi
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|a (IN-ChSCO)ASTMSTP26646S
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| 040 |
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|a ASTM
|c SCOPE
|b eng
|e rda
|d PAU
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| 041 |
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|a eng
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| 100 |
1 |
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|a Baroch, CJ.,
|e author
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| 245 |
1 |
0 |
|a Effect of Irradiation on the Mechanical Properties of 19-9DL Alloy /
|c AL. Lowe, CJ. Baroch
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| 264 |
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1 |
|a West Conshohocken, Pa. :
|b ASTM International,
|c 1970
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| 300 |
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|a 1 online resource (10 pages) :
|b illustrations, figures, tables
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| 336 |
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|a text
|2 rdacontent
|b txt
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| 337 |
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|a computer
|2 rdamedia
|b c
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| 338 |
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|a online resource
|2 rdacarrier
|b cr
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| 347 |
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|a text file
|b PDF
|2 rda
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| 504 |
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|a Includes bibliographical references
|b 5
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| 506 |
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|a Restricted for use by site license.
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|a Austenitic stainless steels, which are used extensively in water cooled nuclear reactors, are presently being considered for use in fast reactors; however, these steels are susceptible to irradiation embrittlement and irradiation induced swelling. The modified austenitic stainless steel 19-9DL alloy, on the other hand, exhibits good creep strength at high temperatures. Babcock & Wilcox conducted an exploratory program to determine the effects of irradiation on 19-9DL alloy at temperatures of 55 C (130 F), 343 C (650 F), and 413 C (775 F). The maximum fluence in this program (9.2x1020 n/cm2, E>1 MeV) was too low to prove the suitability of the alloy's potential for fast reactor applications but did show that additional evaluation of the alloy for fast reactors was warranted. The results indicate that the alloy is suitable for use in thermal reactors, since the ductility at 413 C was greater than 10 percent at a fluence approaching 1x1021 n/cm2, E>1 MeV
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| 541 |
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|a ASTM International
|3 PDF
|c Purchase price
|h USD25
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| 588 |
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|a Description based on publisher's website, viewed February 19, 2016
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| 650 |
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|a Annealing
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| 650 |
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|a Austenitic stainless steels
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| 650 |
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|a Bituminous materials
|x Testing
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| 650 |
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|a Creep properties
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| 650 |
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0 |
|a Ductility
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| 650 |
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|a Elongation
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| 650 |
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0 |
|a Fast reactors (nuclear)
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| 650 |
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|a Irradiation
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| 650 |
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0 |
|a Mechanical properties
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| 650 |
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|a Neutron irradiation
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| 650 |
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|a Nuclear fuel cladding
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| 650 |
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|a Solutions, Solid
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| 650 |
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|a Thermal reactors
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| 650 |
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|a Ultimate strength
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| 650 |
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|a Water cooled reactors
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| 650 |
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|a Yield strength
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| 650 |
1 |
4 |
|a Irradiation
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| 650 |
2 |
4 |
|a Annealing
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| 650 |
2 |
4 |
|a Austenitic stainless steels
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| 650 |
2 |
4 |
|a Creep properties
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| 650 |
2 |
4 |
|a Ductility
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| 650 |
2 |
4 |
|a Elongation
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| 650 |
2 |
4 |
|a Fast reactors (nuclear)
|
| 650 |
2 |
4 |
|a Mechanical properties
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| 650 |
2 |
4 |
|a Neutron irradiation
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| 650 |
2 |
4 |
|a Nuclear fuel cladding
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| 650 |
2 |
4 |
|a Solid solutions
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| 650 |
2 |
4 |
|a Thermal reactors
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| 650 |
2 |
4 |
|a Ultimate strength
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| 650 |
2 |
4 |
|a Water cooled reactors
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| 650 |
2 |
4 |
|a Yield strength
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| 700 |
1 |
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|a Lowe, Al,
|e author
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| 710 |
2 |
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|a ASTM International
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| 710 |
2 |
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|a American Society for Testing and Materials
|t Selected Technical Papers.
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| 710 |
2 |
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|a American Society for Testing and Materials
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| 740 |
0 |
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|a ASTM digital library
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| 999 |
1 |
0 |
|i facc6a71-657c-40e7-ae0b-099b52ff86c1
|l 9978997863503681
|s US-PU
|m effect_of_irradiation_on_the_mechanical_properties_of_19_9dl_alloy_________1970_______astmia________________________________________baroch__cj_________________________e
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