Effects of Loading Conditions and Types of Motion on PWR Fuel Rod Cladding Wear

Joulin, T. P.; ́rout, F. M. Gue; Lina, A.; Moinereau, Dominique

doi:10.1115/imece2002-32837

Cited by 11 publications

(7 citation statements)

References 2 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…Selected cladding wear data from the literature are listed in Table 8. A higher pre-load in the AECL assembly seems to decrease the wear coefficient because of the reduced relative motion between the grid and rod [18]. In contrast, in the current work, a higher load increases the wear coefficient of a cladding in as-received condition when rubbed against either the springs or the dimples.…”

Section: Resultscontrasting

confidence: 56%

See 1 more Smart Citation

Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation

Cooley

Shaw

et al. 2016

Wear

View full text Add to dashboard Cite

In the cores of pressurized water nuclear reactors, water-flow induced vibration is known to cause claddings on the fuel rods to rub against their supporting grids. Such grid-to-rod-fretting (GTRF) may lead to fretting wear-through and the leakage of radioactive species. The surfaces of actual zirconium alloy claddings in a reactor are inevitably oxidized in the high-temperature pressurized water, and some claddings are even pre-oxidized. As a result, the wear process of the surface oxide film is expected to be quite different from the zirconium alloy substrate. This study attempts to measure the wear coefficients of zirconium claddings without and with pre-oxidation rubbing against grid samples using a bench-scale fretting tribometer. Results suggest that the volumetric wear coefficient of the pre-oxidized cladding is 50 to 200 times lower than that of the untreated cladding. In terms of the linear rate of wear depth, the pre-oxidized alloy wears about 15 times more slowly than the untreated cladding. Fitted with the experimentally-determined wear rates, a stage-wise GTRF engineering wear model demonstrates good agreement with inreactor experience in predicting the trend of cladding lives.

show abstract

Section: Resultscontrasting

confidence: 56%

“…Pre-oxidized A commercially-made reciprocating test machine (Ducom TM Instruments, Chicago, USA) was modified for fretting tests ( The test parameters of two other GTRF simulators, the VIPER loop test at Westinghouse and the CANFRET test at AECL (Canada), are listed in Table 2, respectively (based on information in [11,18]…”

Section: As-receivedmentioning

confidence: 99%

Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation

Cooley

Shaw

et al. 2016

Wear

View full text Add to dashboard Cite

show abstract

“…Nevertheless, our simulations predict that irradiation growth and creep have a non-negligible contribution to the gap opening. While there is not much quantitative data available in the literature regarding gap opening during in-reactor conditions, out-of-reactor experiments on fuel rod assemblies generally use a transverse displacement amplitude in the range 5-30 μm for grid-to-rod fretting wear tests [7], [43]. In another analytical study [18], a grid-to-rod gap of 10 μm was predicted at end of life for Zircaloy-4 spacer grid and clad, with consideration for irradiation growth and creep using phenomenological constitutive models.…”

Section: Evolution Of Contact Openingmentioning

confidence: 99%

“…Research has been directed at understanding the involved phenomena and alleviating the root causes responsible for GTR fretting wear. Out-of-reactor experiments study the effect on the wear characteristics of coolant flow induced vibrations and pressure fluctuations [7]- [9], of spring and dimple geometries [10]- [12], and of pre-sized gaps between the spacer grid and the fuel rods [9], [13]. In these experiments, the gaps between the springs and dimples and the fuel rod are generally pre-set to account for the effect of irradiation-induced contact relaxation and cladding creep at the end of life (cf.…”

Section: Introductionmentioning

confidence: 99%

Finite element simulation of gap opening between cladding tube and spacer grid in a fuel rod assembly using crystallographic models of irradiation growth and creep

Patra

Tomé

2017

Nuclear Engineering and Design

View full text Add to dashboard Cite

A physically-based crystal plasticity framework for modeling irradiation growth and creep is interfaced with the finite element code ABAQUS to study the contact forces and the gap evolution between the spacer grid and the cladding tube as a function of irradiation in a representative section of a fuel rod assembly. Deformation mechanisms governing the gap opening are identified and correlated to the texture-dependent material response. In the absence of coolant flow-induced vibrations, these simulations predict the contribution of irradiation growth and creep to the gap opening between the cladding tube and the springs and dimples on the spacer grid. The simulated contact forces on the springs and dimples are compared to available experimental and modeling data. Various combinations of external loads are applied on the springs and dimples to simulate fuel rods in the interior and at the periphery of the fuel rod assembly. It is found that loading conditions representative (to a first order approximation) of fuel rods at the periphery show higher gap opening. This is in agreement with in-reactor data, where rod leakages due to the synergistic effects of gap opening and coolant flow-induced vibrations were generally found to occur at the periphery of the fuel rod assembly.

show abstract

“…During the nuclear component modeling, one of these parameters i.e. wear volume parameter is used that led to the development of ''work rate model'' which in turn is used to find out the wear in the tubes that are used in nuclear steam generator [4] and nuclear fuel tubes [5]. This is given as Eq.…”

Section: Introductionmentioning

confidence: 99%

Wear Analysis of Tube-Baffle Vibration Interaction in a Tube Bundle

et al. 2019

View full text Add to dashboard Cite

Wear analysis has been carried out on aluminum tube in parallel triangular tube bundle with P/D ratio of 1.45 and the baffle plate placed at the mid-span of the bundle. The cross-flow velocity of water ranges from 0.18 m/s to 0.55 m/s. For each flow velocity, tests were performed on five different flow velocities and three-time spans (tube-baffle interaction) i.e., 30, 60, and 90 min. The experiments were designed in a unique way to study the relationship between cross flow velocity and tube baffle interaction time with wear produced on the tube surface. This is the first study of its kind in which wear analysis of a flexible tube has been done in the water tunnel to simulate the real environment of the heat exchanger tubes. The vibration response of the monitored tube was acquired using a tri-axial wireless accelerometer. A Scanning Electron Microscope (SEM) was used for the measurement of wear sizes. It was observed that the transverse amplitude rises significantly at higher velocities as compared to streamwise amplitude results in significant wear in the transverse direction. The wear sizes also tend to increase with the cross-flow velocity and become prominent at higher velocities. A similar trend has been observed for the wear area and wears volume. Also with the increase in the tube-baffle interaction span, the wear scar tends to be deeper and wider significantly at higher velocities.INDEX TERMS Flow-induced vibrations, fretting wear, parallel triangular, wear scar.

show abstract

Effects of Loading Conditions and Types of Motion on PWR Fuel Rod Cladding Wear

Cited by 11 publications

References 2 publications

Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation

Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation

Finite element simulation of gap opening between cladding tube and spacer grid in a fuel rod assembly using crystallographic models of irradiation growth and creep

Wear Analysis of Tube-Baffle Vibration Interaction in a Tube Bundle

Contact Info

Product

Resources

About