Wear of zirconium alloys due to fretting and periodic impacting

doi:10.1016/0043-1648(79)90167-4

Wear

Volume 55, Issue 2, August 1979, Pages 369-384

https://doi.org/10.1016/0043-1648(79)90167-4 Get rights and content

Abstract

Impact fretting of zirconium alloys was studied in demineralized water at temperatures up to 97 °C. Results showing the effects of motion parameters, test duration, material combination and temperature (23–97 °C) are presented. The correlation between the electrical contact resistance and wear is good for zirconium alloy combinations. Continuous monitoring of the change in electrical contact resistance helped to reveal the wear process during tests.

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As a safe, clean, efficient and low-carbon energy, nuclear power is gaining increasing interest in global energy development. However, working in the extremely harsh environment of high temperature, high pressure and high-speed flowing coolant for a long time, fuel assemblies in nuclear reactors face the threat of radioactivity leakage due to grid-to-rod fretting (GTRF) damages. GTRF failures are caused by fuel assembly vibrations induced by turbulent and heterogeneous flow, which are always present, particularly in the entrance region of fuel assemblies. In addition, with the accumulation of burnup, the loss of contact and the appearance of gap in the grid-rod system due to irradiation creep, swelling and growth could significantly increase the GTRF damage. Since fuel rod failures caused by GTRF was observed in 1960s, continuous efforts have been paid to investigate and mitigate GTRF issues, and a lot of academic achievements and engineering progresses have been accumulated. However, in current operating nuclear power plants, GTRF still accounts for more than 50% of all the causes of fuel rod failures. Therefore, a comprehensive and systematic understanding based on previous scientific research and engineering experience is of great value in mitigating GTRF issues. In this paper, an extensive review of GTRF issue is conducted from the perspectives of historical background, driving mechanisms, methodologies-results, and potential mitigation measures in engineering applications. The current status of research on GTRF, including phenomenological mechanisms, affecting factors, research methodologies, as well as impacts of the latest innovative material technologies on GTRF, are highlighted and discussed.
Experimental research progress of grid to rod fretting in pressurized water reactors
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Fuel rods are the most important safety barrier in pressurized water reactors (PWR) and operates at high temperature, high pressure, high speed, and strong irradiation water condition. The grid-to-rod fretting (GTRF) induced by flow-induced vibration (FIV) is the main factor that causes failures in fuel rods. The effects of fretting parameter, environmental conditions, irradiation, fretting wear mode, and spacer grid on the GTRF are systematically reviewed, and the key influencing factor on GTRF is analyzed. The accident tolerant fuel (ATF) coating and its tribological behavior are investigated. Finally, key research guidance and popular topics consideration GTRF are proposed.
An approach to grid-to-rod fretting wear modeling using dynamic mortar contact
2023, Progress in Nuclear Energy
This article presents the development of a grid-to-rod fretting (GTRF) modeling methodology that utilizes dynamic mortar contact. We leverage a recently developed computational framework for modeling thermomechanical contact in dynamic simulations from the nuclear fuel performance code Bison. Usage of mortar contact ensures a smooth distribution of normal and frictional contact forces, displacements, and velocities on the contact interface, thus facilitating an accurate computation of wear. The integration within an advanced nuclear fuel performance code enables analysis of complex interactions between the fuel, cladding, and spacer grid. Such interactions include swelling, creep, fuel fragmentation, burnup, plenum pressure, and fission gas release. Our methodology is demonstrated via the modeling of two spacer grid geometries and the influence of fuel-cladding mechanical contact on the generation of fretting wear.
Mechanism of wear in zircaloy-4 under different loading conditions
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Core components of nuclear reactor are generally made of zirconium alloys and some of them are prone to low amplitude vibration leading to material wear at contact interface. Reciprocating wear tests were conducted between zircaloy-4 (Zr-4) under dry and water-submerged conditions. To confirm the material transfer due to adhesion, wear tests were also conducted between Zr-4 and stainless steel. To understand the wear mechanism, worn surfaces were analyzed using scanning electron microscopy (SEM), field emission scanning electron microscopy (FESEM), energy dispersive spectroscopy (EDS), Raman spectroscopy and X-ray photoelectron spectroscopy (XPS). Coefficient of friction (COF) increases with increasing fretting duration and decreases with increasing frequency and displacement amplitude. Greater COF is observed under dry condition than that under submerged condition. The wear mechanism changes from adhesion to abrasion with increasing displacement amplitude.
Understanding the influence of graphene-based lubricant/coating during fretting wear of zircaloy
2023, Wear
Citation Excerpt :
Despite this, still lacking is comprehensive investigations and understandings related to the effectiveness of different lubricants and/or surface coatings towards reducing the friction/wear damages of the same. Ko [7] has reviewed some aspects associated with the reciprocatory wear of zircaloy. Seemingly, a threshold amplitude exists at about 200 μm tangential motion, above which fretting induced wear damage increases drastically, in addition to increasing with normal load.
In order to study the relative effectiveness of oil-based and graphene-based lubrications towards suppressing the friction and wear damage of zircaloy-4 during low amplitude linear reciprocatory wear, fretting wear tests were performed with varying normal loads (5–20 N) under unlubricated, oil-lubricated (using Castrol TDN-81) and graphene-lubricated/coated (using ∼7 layers graphene film) conditions against stainless steel ball. While the oil-based lubricant has been found to be more effective towards lowering the co-efficient of friction at lower loads, the effectiveness of the graphene-based lubricant/coating increase significantly with an increase in load. In the contexts of suppressing the extent/severity of wear damage, the graphene-based lubricant/coating has been found to be significantly more effective than the oil-based lubricant. In fact, increase in effectiveness of the graphene-based lubricant/coating with load results in the wear volume remaining nearly invariant upon increase in load; thus, causing reduction in wear rate. This is in contrast to the increase in wear volume/rate with increase in load for the unlubricated and oil lubricated conditions; such that the wear volume and rate are lower for the graphene-lubricated/coated condition by more than a factor of 2 with respect to the oil-lubricated condition at 20 N. In the unlubricated condition severe sub-surface cracking induced material removal/delamination, presumably aided by incorporation of Fe and O due to tribochemical reaction, form the major mechanisms of wear damage of zircaloy-4. By contrast, all these get suppressed in the presence of the graphene-based lubricant/coating, even at 20 N; thus, leading to significant reduction in the overall wear damage.
Nuclear power systems
2022, Fretting Wear and Fretting Fatigue: Fundamental Principles and Applications
This chapter provides an overview of the fretting wear damage in nuclear power systems and its economic and safety impact. Description of the main critical safety components of the PWR and CANDU reactors that are susceptible to fretting damage are identified, mainly, the steam generator and the reactor fuel channel and fuel assembly. A system approach to the nonlinear relationships and interactions between forcing functions and fretting damage is presented and a methodology for predicting the long-term fretting damage of nuclear structural components due to flow-induced vibrations (FIV) and random turbulent excitation is discussed. A critical review and analysis of various wear energy parameters is presented. The chapter is focused on the effect of various process parameters, e.g., mode of excitation, impact-sliding motion, contact configuration/support geometry, material combinations, temperature, water chemistry, surface treatment, and the tribo-oxide formation on the severity of the damage. The chapter is concluded with the identification of gaps in the current state of knowledge and the recommendations for future work.

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^☆: Presented at the International Conference on Wear of Materials, 1979, Dearborn, Michigan, April 16 –18, 1979.

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Wear of zirconium alloys due to fretting and periodic impacting☆

Abstract

Wear

Wear

Dynamics of fuel strings in axial flow

Ann. Nucl. Energy

An analytical and experimental investigation of the vibration of cylindrical reactor fuel elements in two-phase parallel flow

Nucl. Sci. Eng.

Flow-induced vibration of nuclear power station components

Mechanism of fretting corrosion

J. Appl. Mech.