Vibration analysis of PWR fuel rod

Kim, Hyeong Koo; Kim, Moon Saeng

doi:10.1016/j.jsv.2004.04.024

Cited by 13 publications

(5 citation statements)

References 7 publications

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“…7, it is in the range of the natural frequencies based on the above two slenderness ratios. Because there is a gap between the cylinders in the position where the wire is crossing, the experiment value is closer to the frequency calculated in the case 3, although the turbulence caused by wire spacer has an ef-fect on the static deformation, the dynamic response is similar to the bare fuel rod [27]. And thus the calculated results are reasonable.…”

Section: Analysis Of Engineering Examplesupporting

confidence: 70%

Dynamic Characteristics Analysis of Cylinders Bundle Coupling System Under Axial Flow

Shu

Yang

et al. 2021

mech

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Nuclear reactor fuel assemblies are mainly composed of cylinders bundle(CB), calculating the dynamics characteristics of CB under axial flow can lay a foundation for predicting fretting wear and vibration fatigue. In the paper, the CB coupling dynamic model of forced vibration under pulsating flow is established. And the stability analysis and natural frequency calculation of the CB system under steady flow are compared with the existing results to verify the model. Finally, the Runge-Kutta method is applied to solve the forced vibration equation of the CB under pulsating flow. The influence of the pulsating parameters m, w0, on the amplitude-frequency characteristics and the motion trajectory of the midspan cross section of the CB under forced vibration are analyzed and discussed. The results show that the pulsating parameters have an important influence on the vibration of the CB system.

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Section: Analysis Of Engineering Examplesupporting

confidence: 70%

Dynamic Characteristics Analysis of Cylinders Bundle Coupling System Under Axial Flow

Shu

Yang

et al. 2021

mech

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“…At the peak oxide location, the region of enriched hydrogen content can extend more than 100 μm (3·94 mil) from the outer surface 25. For fuel rods discharged with burnup of 45–50 GWd MTU −1 , the amount of hydrogen collected at the outer surface exceeds the saturation limit, and as much as 120 ppm hydrogen is available for hydride formation in the form of radial hydrides or hydride blisters23 when the cladding tubes are cooled from 350 to 45°C (662–113°F). Because the fracture toughness of zirconium hydride [1–3 MPa m −1/2 (0·91–2·73 ksi in.…”

Section: Crack Size Distributions In Zirconium Claddingmentioning

confidence: 99%

“…1/2 )] is slightly less than that of zirconium oxide [≧2·6 MPa m −1/2 (2·37 ksi in. 1/2 )], cracking of the oxide layer at the maximum thickness location is likely to be followed by fracture of the hydride layer or hydride blister below the oxide layer 23. Under this circumstance, the maximum crack length in the fuel rod cladding would be on the order of 200 μm (7·87 mil).…”

Section: Crack Size Distributions In Zirconium Claddingmentioning

confidence: 99%

“…Under this circumstance, the maximum crack length in the fuel rod cladding would be on the order of 200 μm (7·87 mil). Its subsequent growth by DHC if it occurs could lead to rod failure 23. Detailed descriptions of cladding failure resulting from crack initiation at the outside wall are available in the literature 21,25,27,36.…”

Section: Crack Size Distributions In Zirconium Claddingmentioning

confidence: 99%

“…The causes and mitigation methods for fuel failure in BWRs were reviewed in an IAEA Technical Meeting 18. Kim23 reported that primary and secondary hydriding in Zircaloy-2 cladding played critical roles in recent failures of nuclear fuel rods in Korean BWR power plants. The roles of hydrides in these rod failures are similar to those reported by several investigators,16,19 – 22 as summarised earlier in the section on ‘Introduction’.…”

Section: Crack Size Distributions In Zirconium Claddingmentioning

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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Towards modeling of Grid to Rod and the effect of its nonlinear dynamic behaviors on grid-to-rod fretting wear

Huang

YANG

2023

Annals of Nuclear Energy

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Vibration analysis of PWR fuel rod

Cited by 13 publications

References 7 publications

Dynamic Characteristics Analysis of Cylinders Bundle Coupling System Under Axial Flow

Dynamic Characteristics Analysis of Cylinders Bundle Coupling System Under Axial Flow

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

Towards modeling of Grid to Rod and the effect of its nonlinear dynamic behaviors on grid-to-rod fretting wear

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