Characteristics of Axial Splits in Failed BWR Fuel Rods

Lysell, G.; Grigoriev, V.

doi:10.1002/9781118787618.ch123

Cited by 10 publications

(14 citation statements)

References 1 publication

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“…Irradiation appears to exert only a small effect on the K IH value of Zr-2?5Nb tested at 130-350uC (266-662uF). Sagat et al 43 reported a small decrease in K IH as the fluence increased to about 1610 25 n m 22 (1?086 10 26 n ft 22 ) for fast neutrons with energy above 1 MeV at an irradiation temperature in the range of 250-290uC (482-554uF), but no increase upon further increase in fluence up to 8610 25 n m 22 (8?61610 26 n ft 22 ), as shown in Fig. 13.…”

Section: Irradiation Effectsmentioning

confidence: 82%

“…22,37,38,44,45 In both cases, the length of individual hydrides was about 10-15 mm (0?394-0?59 mil). 22,38 These experimental observations suggested that a critical hydride length of 10-20 mm (0?394-0?79 mil) may apply to both Zr-2?5Nb and Zircaloy-2. In addition, the critical hydride length [10-20 mm (0?394-0?787 mil)] is essentially identical to the a o value.…”

Section: Threshold Stress Intensity Factor K Ihmentioning

confidence: 88%

“…Sagat el al. 43 also measured the DHC velocity and K IH in Zr-2?5Nb that were irradiated at temperatures in the range of 250-290uC (482-554uF) in fast neutrons with a high energy (E.1 MeV) and fluxes between 1?6610 17 and 1?8610 18 n m 22 s 21 (1?72610 18 and 1?94610 19 n ft 22 s 21 ) to fluences between 0?016 10 25 and 9?8610 25 n m 22 (1?08610 24 and 1?066 10 27 n ft 22 ). Their results indicated that these levels of neutron irradiation reduced K IH by about 20% 43 and increased the DHC velocity by a factor of five.…”

Section: Effects Of Irradiation On Dhcmentioning

confidence: 99%

“…[16][17][18] Some failures of Zircaloy-2 cladding tubes with and without a Zr lined barrier were reported during service in BWRs. 16,[19][20][21][22][23][24] From 1989 to 1993, 16,[19][20][21] almost 30% of the failed Zr lined rods exhibited long axial cracks that exceeded 15 cm in length. Some of these cracks were secondary defects, while a few were primary defects.…”

Section: Introductionmentioning

confidence: 99%

“…For high burnup fuel rods, Pierron et al 35 recently showed evidence of oxide cracking on the outer surface of an irradiated Zircaloy-4 tube operated to a fuel burnup of 67 GWd MTU 21 and with fast neutron fluence of 1?3610 22 n cm 22 . A solid hydride rim was also observed to lie beneath the oxide layer.…”

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: Irradiation Effectsmentioning

confidence: 82%

Section: Threshold Stress Intensity Factor K Ihmentioning

confidence: 88%

Section: Effects Of Irradiation On Dhcmentioning

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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Hydride Reorientation and Delayed Hydride Cracking of Spent Fuel Rods in Dry Storage

Kim

2009

Metall Mater Trans A

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The aim of this work is to investigate the effect of thermal creep during vacuum drying of spent fuel rods on hydride reorientation and their delayed hydride cracking (DHC) susceptibility. To these ends, we analyzed Tsai's thermal creep results of irradiated Zircaloy-4 cladding segments from two pressurized water reactors and Simpson and Ells' observation where zirconium alloy cladding tube failed during long-term storage at room temperature. On cooling under 190 MPa, the spent fuel rods crept to 3.5 pct strain during vacuum drying showed large radial hydrides, while the ones crept to 0.35 pct strain had very fine radial hydrides. Thus, it is suggested that prior creep deformation promotes hydride reorientation in spent fuel rods on cooling after vacuum drying. Evidence for this suggestion is provided by a model experiment. Considering Kim's DHC model and experimental facts showing precipitation of hydrides even at room temperature at stress raisers, we suggest that spent fuel rods would fail by DHC in dry storage if stress raisers are present inside the cladding on cooling to below 180°C, and then axial splits of the failed spent fuel rods would occur by DHC due to fuel expansion by UO 2 oxidation.

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The Effect of Microstructure on Delayed Hydride Cracking Behavior of Zircaloy-4 Fuel Cladding—An International Atomic Energy Agency Coordinated Research Program

Coleman

Grigoriev

Inozemtsev

et al. 2010

Journal of ASTM International

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The rate of delayed hydride cracking (DHC) has been measured in Zircaloy-4 fuel cladding in several metallurgical conditions using the pin-loading tension technique. In light water reactor (LWR) cladding in the cold-worked and cold-worked and stress-relieved conditions, the cracking rate followed Arrhenius behavior up to about 280 °C, but at higher temperatures the rate declined with no cracking above 300°C. Non-LWR cladding appeared to behave in the same manner. In LWR cladding in the recrystallized condition, the cracking rate was highly variable because it depended on KI within the test range up to 25 MPa√m, whereas in the other LWR claddings, cracking rate was independent of KI, indicating that KIH was below 11 MPa√m. The main role of microstructure was to control the material strength; the cracking rate increased as the strength increased. Although all the claddings had a radial texture, it did not protect the cladding from DHC. The DHC fracture surface consisted of flat broken hydrides, often in arcs, but no striations were observed, except in one specimen subjected to thermal cycles.

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Characteristics of Axial Splits in Failed BWR Fuel Rods

Cited by 10 publications

References 1 publication

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

Hydride Reorientation and Delayed Hydride Cracking of Spent Fuel Rods in Dry Storage

The Effect of Microstructure on Delayed Hydride Cracking Behavior of Zircaloy-4 Fuel Cladding—An International Atomic Energy Agency Coordinated Research Program

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