Review of behavior of mixed-oxide fuel elements in extended overpower transient tests in EBR-II [electronic resource]
From a series of five tests conducted in EBR-II, a substantial data base has been established on the performance of mixed-oxide fuel elements in a liquid-metal-cooled reactor under slow-ramp transient overpower conditions. Each test contained 19 preirradiated fuel elements with varying design and pr...
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Online Access |
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| Corporate Author: | |
| Format: | Government Document Electronic eBook |
| Language: | English |
| Published: |
Washington, D.C. : Oak Ridge, Tenn. :
United States. Dept. of Energy ; distributed by the Office of Scientific and Technical Information, U.S. Dept. of Energy,
1994.
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| Subjects: |
| Summary: | From a series of five tests conducted in EBR-II, a substantial data base has been established on the performance of mixed-oxide fuel elements in a liquid-metal-cooled reactor under slow-ramp transient overpower conditions. Each test contained 19 preirradiated fuel elements with varying design and prior operating histories. Elements with aggressive design features, such as high fuel smear density and/or thin cladding, were included to accentuate transient effects. The ramp rates were either 0.1 or 10% ΔP/P/s and the overpowers ranged between ≈60 and 100% of the elements̀ prior power ratings. Six elements breached during the tests, all with aggressive design parameters. The other elements, including all those with moderate design features for the reference or advanced long-life drivers for PNC̀s prototype fast reactor Monju, maintained their cladding integrity during the tests. Posttest examination results indicated that fuel/cladding mechanical interaction (FCMI) was the most significant mechanism causing the cladding strain and breach. In contrast, pressure loading from the fission gas in the element plenum was less important, even in high-burnup elements. During an overpower transient, FCMI arises from fuel/cladding differential thermal expansion, transient fuel swelling, and, significantly, the gas pressure in the sealed central cavity of elements with substantial centerline fuel melting. Fuel performance data from these tests, including cladding breaching margin and transient cladding strain, are correlatable with fuel-element design and operating parameters. These correlations are being incorporated into fuel-element behavior codes. At the two tested ramp rates, fuel element behavior appears to be insensitive to transient ramp rate and there appears to be no particular vulnerability to slow ramp transients as previously perceived. |
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| Item Description: | Published through the Information Bridge: DOE Scientific and Technical Information. 10/01/1994. "anl/et/cp--83008" " conf-950426--5" "DE95013682" 3. international conference on nuclear engineering, Kyoto (Japan), 23-27 Apr 1995. Neimark, L.A.; Tsai, H.; Nagai, S.; Nakae, N. |
| Physical Description: | 13 p. : digital, PDF file. |