Cladding Evaluation in the Yucca Mountain Repository Performance Assessment

Siegmann, E.R.; McCoy, J. Kevin; Howard, Robert

doi:10.1557/proc-608-3

Cited by 2 publications

(6 citation statements)

References 5 publications

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“…The peak cladding temperature history for the centre rod in the waste package during the operation period before disposal is about 325uC (617uF). 49 Because of the high DHC propagation rate near 318uC (604uF), the time to leakage is on the order of a few hours if K.K IH in any cladding tubes. DHC may occur based on internal gas pressure alone if large hydride blisters or radial hydrides are present in the microstructure, the fracturing of which leads to K.K IH .…”

Section: Dhc In Zirconium Alloy Cladding Tubesmentioning

confidence: 99%

“…The cladding stress under normal in-reactor conditions arises mainly from the internal gas pressure inside the cladding tubes. While the gas pressure increases with burnup and fission gas release fraction, there is a general consensus that the cladding stress due to internal gas pressure is sufficiently low that the growth threshold for DHC K IH would not be exceeded unless the cladding tube contained a deep crack 27,47,49,50. On the other hand, the transient stresses encountered in the cladding resulting from interaction with the fuel during in-reactor power ramps can be sufficiently high that they are likely to instigate DHC and could well lead to cladding failure under certain circumstances.…”

Section: Cladding Stressesmentioning

confidence: 99%

“…These gas pressure values correspond to hoop stresses at 27uC (81uF) for 38?4 MPa (5?6 ksi) and 61?8 MPa (9?0 ksi) for the Westinghouse 17617 Lopar (W1717WL) fuel assembly. Using a crack size distribution based on the manufacturing defect size distribution, Siegmann et al [48][49][50] and CRWMS M&O 27 computed the stress intensity factors acting on cracks in cladding tubes in the range of 0?5-2 MPa m 1/2 (0?46-1?82 ksi in. 1/2 ), 27,[48][49][50] which is less than the crack growth threshold K IH for DHC.…”

Section: Introductionmentioning

confidence: 99%

“…Using a crack size distribution based on the manufacturing defect size distribution, Siegmann et al [48][49][50] and CRWMS M&O 27 computed the stress intensity factors acting on cracks in cladding tubes in the range of 0?5-2 MPa m 1/2 (0?46-1?82 ksi in. 1/2 ), 27,[48][49][50] which is less than the crack growth threshold K IH for DHC. Thus, DHC in cladding tubes in the potential repository was considered unlikely by DOE, 27,[48][49][50] which is consistent with assessments made by Cunningham et al 51 Cunningham et al 51 concluded that DHC would not be expected to occur, because the operating stress intensity factors on cracks in cladding tubes are expected to be lower than the crack growth threshold for DHC.…”

Section: Introductionmentioning

confidence: 99%

“…1/2 ), 27,[48][49][50] which is less than the crack growth threshold K IH for DHC. Thus, DHC in cladding tubes in the potential repository was considered unlikely by DOE, 27,[48][49][50] which is consistent with assessments made by Cunningham et al 51 Cunningham et al 51 concluded that DHC would not be expected to occur, because the operating stress intensity factors on cracks in cladding tubes are expected to be lower than the crack growth threshold for DHC. The previous studies, however, did not consider the possibility that the hydride microstructure of cladding tubes could be altered by creep at dry storage 52 or degraded due to the formation of hydride blisters, hydride rims and radially oriented hydrides as the result of high PCMI stresses during power ramps.…”

Section: Introductionmentioning

confidence: 99%

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An assessment of delayed hydride cracking in zirconium alloy cladding tubes under stress transients

Chan

2013

International Materials Reviews

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Zirconium alloy cladding tubes used in nuclear fuel rods are susceptible to delayed hydride cracking, which is a time dependent crack growth process resulting from the stress assisted diffusion of hydrogen to the crack tip, followed by the formation of radial hydrides and the subsequent fracture of the hydrides in the crack tip region. This article reviews the current understanding of the delayed hydride cracking behaviour of zirconium alloy cladding tubes for fuel rods, focusing on the degradation mechanisms in high burnup fuel rods and transient loading scenarios, which could potentially lead to substantial changes in the hydride microstructure and cladding failure by delayed hydride cracking following removal from the reactor and during storage and disposal of spent nuclear fuel rods in a waste repository. A brief summary of the general characteristics of delayed hydride cracking in zirconium alloy cladding is presented first. Relevant information on the cladding stresses under various usage conditions is then compiled and categorised into several characteristic stress transients that can be anticipated during reactor operation. Delayed hydride cracking in cladding tubes under stress transients is then examined under various temperatures, cooling rates, burnup levels and loading conditions.

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Section: Dhc In Zirconium Alloy Cladding Tubesmentioning

confidence: 99%

Section: Cladding Stressesmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

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An assessment of delayed hydride cracking in zirconium alloy cladding tubes under stress transients

Chan

2013

International Materials Reviews

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Sensitivity Studies of the Effect of Cladding Degradation on TSPA Results

Siegmann

Devonec²

2002

MRS Proc.

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The Commercial Spent Nuclear Fuel (CSNF) cladding directly influences the dose by reducing the release rate of radionuclides compared to bare fuel. A cladding degradation model was introduced in the Total System Performance Assessment – Viability Assessment (TSPA-VA) and has evolved into the TSPA-SR (Sight Recommendation), Rev 00 and most recently the Supplemental Science and Performance Analyses (SSPA) models. TSPAs are large computer models that predict how the radionuclides might escape through the various barriers and migrate through the different geological regions and predict the dose to the critical population. The major components in the cladding degradation model are initial cladding failure, creep, stress corrosion cracking (SCC), localized corrosion, mechanical failures (due to seismic events and rock overburden) and cladding unzipping. In the bare fuel case, the dose is controlled by the failure rate of the waste packages (WPs), the solubility limits for certain radionuclides, and the rate of diffusion of the radionuclides through the cracks in the WPs. The cladding degradation model involves two steps, perforation followed by unzipping. Sensitivity studies have shown that cladding limits dose primarily by preventing dissolution of the UO2 fuel (when the cladding is not perforated). If all the cladding is assumed to be perforated and cladding unzipping occurs, the peak dose would be within 4% of that of bare fuel. The UO2 dissolution rate (the driving mechanism for cladding unzipping) is sufficiently rapid that assuming instant dissolution increases the dose by only 12% from the base case. In the current cladding degradation model, cladding accounts for a reduction in dose for the first 100,000 years of a factor of 16 when compared with bare fuel. This is because of the low initial cladding failure rate (2.1% including SCC and creep) and few failures from localized corrosion (late WP breach and limited water entering the WPs). Failure of the cladding by rock overburden accounts for an increasing trend in dose at times greater than 100,500 years. This mechanism increases the peak dose and delays the peak from approximately 200,000 to 300,000 years. In conclusion, the cladding degradation model, mostly through preventing water from contacting the UO2, reduces the dose predicted in the TSPA analysis.

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Cladding Evaluation in the Yucca Mountain Repository Performance Assessment

Cited by 2 publications

References 5 publications

An assessment of delayed hydride cracking in zirconium alloy cladding tubes under stress transients

An assessment of delayed hydride cracking in zirconium alloy cladding tubes under stress transients

Sensitivity Studies of the Effect of Cladding Degradation on TSPA Results

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