• 한국윤활학회:학술대회논문집
• /
• 한국윤활학회 2002년도 proceedings of the second asia international conference on tribology
• /
• pp.185-186
• /
• 2002

Fretting of fuel rod cladding material, Zircaloy-4 tube, in PWR nuclear power plants must be reduced and avoided. Nowadays the introduction of surface treatments or coatings is expected to be an ideal solution to fretting damage since fretting is closely related to wear. corrosion and fatigue. Therefore. in this study the fretting wear experiment was performed using TiAlN coated Zircaloy-4 tube as the fuel rod cladding and uncoated Zircaloy-4 as on of grids, especially concentrating on the sliding component. Fretting wear resistance of TiAlN coated Zircaloy-4 tubes was improved compared with that of TiN coated tubes and uncoated tubes and fretting wear mechanisms were brittle fracture and plastic flow at lower slip amplitude but severe oxidation and spallation of oxidative layer at higher ship amplitude.

• Nuclear Engineering and Technology
• /
• 제48권2호
• /
• pp.572-594
• /
• 2016

As one of the Gen-IV nuclear energy systems, a sodium-cooled fast reactor (SFR) is being developed at the Korea Atomic Energy Research Institute. As a long-term national research project, advanced radiation resistant oxide dispersion strengthened steel (ARROS) is being developed as an in-core fuel cladding tube material for a SFR in the future. In this paper, the current status of ARROS development is reviewed and its future prospective is discussed.

• Nuclear Engineering and Technology
• /
• 제45권6호
• /
• pp.811-820
• /
• 2013

SiC has been under investigation as a potential cladding for LWR fuel, due to its high melting point and drastically reduced chemical reactivity with liquid water, and steam at high temperatures. As SiC is a brittle material its behavior during the reflood phase of a Loss of Coolant Accident (LOCA) is another important aspect of SiC that must be examined as part of the feasibility assessment for its application to LWR fuel rods. In this study, an experimental assessment of thermal shock performance of a monolithic alpha phase SiC tube was conducted by quenching the material from high temperature (up to $1200^{\circ}C$) into room temperature water. Post-quenching assessment was carried out by a Scanning Electron Microscopy (SEM) image analysis to characterize fractures in the material. This paper assesses the effects of pre-existing pores on SiC cladding brittle fracture and crack development/propagation during the reflood phase. Proper extension of these guidelines to an SiC/SiC ceramic matrix composite (CMC) cladding design is discussed.

• Nuclear Engineering and Technology
• /
• 제55권7호
• /
• pp.2454-2465
• /
• 2023

Film boiling may lead to burnout of the heating element. Even though burnout does not occur, the heating element is subject to deformation because it is not sufficiently strong to withstand external loads. In particular, the ballooning and rupture of a tube under film boiling are important phenomena in the field of nuclear reactor safety. If the tube-type cladding of nuclear fuel ruptures owing to high internal pressure and thermal load, radioactive materials inside the cladding are released to the coolant. Therefore, predicting the ballooning and rupture is important. This study presents numerical simulations to predict the ballooning behavior and rupture time of a horizontal tube at high internal pressure under saturated film boiling. To do so, a multi-step coupled simulation of conjugated film boiling heat transfer and ballooning using creep model is adopted. The numerical methods and models are validated against experimental values. Two different nonuniform heat flux distributions and four different internal pressures are considered. The three-step simulation is enough to obtain a convergent result. However, the single-step simulation also successfully predicts the rupture time. This is because the film boiling heat transfer characteristics are slightly affected by the tube geometry related to creep ballooning.

• Nuclear Engineering and Technology
• /
• 제53권7호
• /
• pp.2312-2322
• /
• 2021

In recent studies, the creep rate of Zircaloy-4, one of the basic property parameters of the nuclear fuel code, has been commonly used with the axial creep model proposed by Rosinger et al. However, in order to calculate the circumferential deformation of the fuel cladding, there is a limitation that a difference occurs depending on the anisotropic coefficients used in deriving the circumferential creep equation by using the axial creep equation. Therefore, in this study, the existing axial creep law and the derived circumferential creep results were analyzed through a circumferential creep test by the internal pressurization method in the isothermal conditions. The circumferential creep deformation was measured through the optical image analysis method, and the results of the experiment were investigated through constructed IDECA (In-situ DEformation Calculation Algorithm based on creep) code. First, preliminary tests were performed in the isotropic β-phase. Subsequently in the anisotropic α-phase, the correlations obtained from a series of circumferential creep tests were compared with the axial creep equation, and optimized anisotropic coefficients were proposed based on the performed circumferential creep results. Finally, the IDECA prediction results using optimized anisotropic coefficients based on creep tests were validated through tube burst tests in transient conditions.

• Tribology and Lubricants
• /
• 제17권6호
• /
• pp.453-458
• /
• 2001

The fretting wear of a nuclear fuel rod it a dangerous phenomenon. In this study, TiN coating was used to reduce the fretting wear of Zircaloy-4 tube, a nuclear fuel rod cladding material. TiN coating is probably one of the molt frequently and successfully used PVD coatings for the mitigation of fretting wear. The fretting tester was designed and manufactured for this experiment. The number of cycles, slip amplitude and normal load were selected as main factors of fretting wear. The results of this research showed that wear volume was improved 1.3∼3.2 times with TiN coating. The worn surfaces were observed by SEM. Wear mechanism at lower slip amplitude was the brittle cracks and rupture of TiN coating. However, adhesive and abrasive wear were mainly observed on most surfaces at higher slip amplitude.

• 방사성폐기물학회지
• /
• 제19권3호
• /
• pp.307-322
• /
• 2021

This study developed an analytical methodology for the mechanical integrity of spent nuclear fuel (SNF) cladding tubes under external pinch loads during transportation, with reference to the failure mode specified in the relevant guidelines. Special consideration was given to the degraded characteristics of SNF during dry storage, including oxide and hydride contents and orientations. The developed framework reflected a composite cladding model of elastic and plastic analysis approaches and correlation equations related to the mechanical parameters. The established models were employed for modeling the finite elements by coding their physical behaviors. A mechanical integrity evaluation of 14 × 14 PWR SNF was performed using this system. To ensure that the damage criteria met the applicable legal requirements, stress-strain analysis results were separated into elastic and plastic regions with the concept of strain energy, considering both normal and hypothetical accident conditions. Probabilistic procedures using Monte Carlo simulations and reliability evaluations were included. The evaluation results showed no probability of damage under the normal conditions, whereas there were small but considerably low probabilities under accident conditions. These results indicate that the proposed approach is a reliable predictor of SNF mechanical integrity.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2004년도 요약집 춘계학술발표대회
• /
• pp.301.2-301.2
• /
• 2004

• 한국방사성폐기물학회:학술대회논문집
• /
• 한국방사성폐기물학회 2016년도 추계학술논문요약집
• /
• pp.193-194
• /
• 2016

• Nuclear Engineering and Technology
• /
• 제52권11호
• /
• pp.2565-2571
• /
• 2020

Background: Understanding the behaviour of nuclear fuel claddings by conducting burst test on single cladding tube under simulated loss-of-coolant accident conditions and developing theoretical cum empirical predictive computer codes have been the focus of several investigations. The developed burst criterion (a) assumes symmetrical deformation of cladding tube in contrast to experimental observation (b) interpolates the properties of Zircaloy-4 cladding in mixed α+β phase (c) does not account for azimuthal temperature variations. In order to overcome all these drawbacks of burst criterion, it is reasoned that artificial intelligence technique may be a better option to predict the burst parameters. Methods: Artificial neural network models based on feedforward backpropagation algorithm with logsig transfer function are developed. Results: Neural network architecture of 2-4-4-3, that is model with two hidden layers having four nodes in each layer is found to be the most suitable. The mean, maximum, and minimum prediction errors for this optimised model are 0.82%, 19.62%, and 0.004%, respectively. Conclusion: The burst stress, burst temperature, and burst strain obtained from burst criterion have average deviation of 19%, 12%, and 53% respectively whereas the developed neural network model predicted these parameters with average deviation of 6%, 2%, and 8%, respectively.