Author: D. F. Sears
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Languages : en
Pages : 0
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Five canflex fuel bundles were irradiated in the u2 loop in nru during 1990. the 43-element bundles contained two element sizes: 11.5 mm diameter elements with 0.33 mm thin-wall zr-4 cladding in the outer and intermediate rings, and 13.5 mm diameter elements with 0.35 mm thin-wall cladding in the inner ring. three bundles contained uo2 enriched to 2.59% u-235 in u, and two contained natural uo2 in the intermediate and outer ring of elements, with 2.59% enriched uo2 in the inner ring. the bundles overheated when, due to loss of class iv electrical power to the main loop pumps, the coolant flow stagnated briefly in the down leg of the u loop, then reversed direction as convective cooling was established. although the bundle power decayed rapidly after the reactor tripped due to the loss of electrical power, the stored energy and decay heat were sufficient to cause overheating of the bundles during the brief stagnation period, and this in turn led to steam generation which resulted in high coolant pressure (sheath temperature in excess of 600 degrees c and coolant pressure above 11.2 mpa). post-irradiation examinations revealed that the sheath in some elements had collapsed into the gaps between the pellet stack and end caps, and between adjacent pellets; however, no defects were attributed to sheath collapse. two of the elements had defects due to hydride blisters, but the source of the hydrogen was not from a breach in the cladding or end caps. the defects were attributed to primary hydriding. the results show that even though the bundles were subjected to abnormally high temperatures and pressures, and were damaged as a result, none of the collapsed elements leaked. this indicates that thin-wall cladding provides a significant margin of safety during transient high temperature and pressure excursions.
Author: B. A. Surette
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Languages : en
Pages : 0
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Author: P. J. Valliant
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Languages : en
Pages : 0
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Zircaloy-4 fuel sheath specimens 15.2 mm diameter by 0.42 mm wall thickness with beryllium-brazed appendages have been laboratory tested at high temperature with internal gas pressurization and inert gas shielding of the outer surfaces. with sufficient hoop stress at temperatures less than 1073 k cracks, which were due to a beryllium assisted crack penetration mechanism, were found to occur at the appendages. formulae to predict the beryllium penetration and rupture have been derived by westinghouse canada ltd. based on data from direct resistance-heated specimens. the formulae are thought to be able to predict the phenomenon in fuel sheaths with changing temperature/pressure histories. a series of tests was done by canadian general electric to determine the formulae ability to predict the occurrence of beryllium assisted crack penetration in specimens heated indirectly with internal heaters. the tests showed: 1) there is an incubation period before any beryllium assisted crack penetration occurs. 2) at sheath temperatures below approximately equal to 1073 k sheath rupture by excessive mechanical strain will preclude rupture by beryllium assisted crack penetration. 3) the formulae predicted rupture times above and below the braze alloy melting temperature within the previously determined confidence limits. 4) predictions of beryllium assisted crack penetration rupture for specimens with changing temperature/pressure histories were of similar accuracy to predictions of isothermal tests. 5) specimens with an appendage plane of one bearing pad and three spacer pads behaved similarly to the specimens with four spacer pads on which the formulae were based.
Author: D. Grant
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Languages : en
Pages : 0
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Author: L. Jansson
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Languages : en
Pages : 0
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Category : Nuclear fuels
Languages : en
Pages : 516
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Author: P. G. Anderson
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Languages : en
Pages : 0
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Author: L. N. Herbert
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Languages : en
Pages : 0
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Author: S. Stroes-Gascoyne
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Category :
Languages : en
Pages : 0
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Author: J. R. Kelm
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Category :
Languages : en
Pages : 0
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