Postirradiation examination of the HT9 clad fuel test X425 at 2.9% burnup [electronic resource]
The X425 experiment was the first EBR-II subassembly to be irradiated with U-Pu-Zr metallic fuel clad in the HT9 alloy. This report summarizes our initial postirradiation examination of selected elements from X425 at 2.9% peak burnup. Fuel microstructure, swelling behavior, fission gas release, and...
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Online Access (via OSTI) |
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| Format: | Government Document Electronic eBook |
| Language: | English |
| Published: |
Washington, D.C. : Oak Ridge, Tenn. :
United States. Department of Energy. ; distributed by the Office of Scientific and Technical Information, U.S. Department of Energy,
1987.
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| Subjects: |
MARC
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| 245 | 0 | 0 | |a Postirradiation examination of the HT9 clad fuel test X425 at 2.9% burnup |h [electronic resource] |
| 260 | |a Washington, D.C. : |b United States. Department of Energy. ; |a Oak Ridge, Tenn. : |b distributed by the Office of Scientific and Technical Information, U.S. Department of Energy, |c 1987. | ||
| 300 | |a 58 p. : |b digital, PDF file. | ||
| 336 | |a text |b txt |2 rdacontent. | ||
| 337 | |a computer |b c |2 rdamedia. | ||
| 338 | |a online resource |b cr |2 rdacarrier. | ||
| 500 | |a Published through SciTech Connect. | ||
| 500 | |a 11/01/1987. | ||
| 500 | |a "anl-ifr--81" | ||
| 500 | |a "TI88025660" | ||
| 500 | |a Pahl, R G; Beck, W N; Sanecki, J E. | ||
| 520 | 3 | |a The X425 experiment was the first EBR-II subassembly to be irradiated with U-Pu-Zr metallic fuel clad in the HT9 alloy. This report summarizes our initial postirradiation examination of selected elements from X425 at 2.9% peak burnup. Fuel microstructure, swelling behavior, fission gas release, and fuel/clad chemical interaction are discussed. | |
| 536 | |b W-31109-ENG-38. | ||
| 650 | 7 | |a Alloy-Ht-9. |2 local. | |
| 650 | 7 | |a Physical Radiation Effects. |2 local. | |
| 650 | 7 | |a Fuel Cans. |2 local. | |
| 650 | 7 | |a Post-Irradiation Examination. |2 local. | |
| 650 | 7 | |a Burnup. |2 local. | |
| 650 | 7 | |a Microstructure. |2 local. | |
| 650 | 7 | |a Nuclear Fuels. |2 local. | |
| 650 | 7 | |a Swelling. |2 local. | |
| 650 | 7 | |a Uranium. |2 local. | |
| 650 | 7 | |a Plutonium. |2 local. | |
| 650 | 7 | |a Zirconium. |2 local. | |
| 650 | 7 | |a Materials Testing. |2 local. | |
| 650 | 7 | |a Reactor Materials. |2 local. | |
| 650 | 7 | |a General Studies Of Nuclear Reactors. |2 edbsc. | |
| 650 | 7 | |a Engineering. |2 edbsc. | |
| 650 | 7 | |a Materials Science. |2 edbsc. | |
| 710 | 2 | |a Argonne National Laboratory. |4 res. | |
| 710 | 1 | |a United States. |b Department of Energy. |4 spn. | |
| 710 | 1 | |a United States. |b Department of Energy. |b Office of Scientific and Technical Information. |4 dst. | |
| 856 | 4 | 0 | |u http://www.osti.gov/scitech/biblio/713968 |z Online Access (via OSTI) |
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