Jung-Min Suh

Impact of Nuclear Fuel Assembly Design on Grid-to-Rod Fretting Wear

The fluid-induced vibration at the nuclear fuel assembly may cause grid-to-rod fretting wear, and subsequently, fuel rod perforation. The fluid-induced grid-to-rod fretting wear occurs with a certain fuel assembly design and/or in certain nuclear power plants, which may be strongly correlated with external and internal vibration causes acting on the fuel assembly. The external vibration causes may include reactor coolant flow velocity, nonuniform radial flow profiles caused by reactor internals, and interfuel assembly gap and fuel assembly-shroud gap, whereas the internal vibration causes may include asymmetric mixing vane pattern across the spacer grid assembly and inadequate spacer grid design. In this study, the external and internal vibration causes are described including three internal vibration mechanisms acting on the fuel assembly. The impact of internal vibration on the grid-to-rod fretting wear was investigated, based on out-of-pile vibration and grid-to-rod fretting wear test results for two kinds of fuel assembly designs as well as their operating experiences in commercial reactors. In addition, fuel assembly design optimization guidelines are proposed to eliminate the grid-to-rod fretting wear-induced failure.

STRUCTURAL INTEGRITY EVALUATION OF NUCLEAR FUEL WITH REDUCED WELDING CONDITIONS

Welding is required for a connection between two different components in the nuclear fuel of a pressurized water reactor. This work relies on a mechanical experiment and analytic results to investigate the structural integrity of nuclear fuel in a situation where some components are not welded to each other. A series of lateral vibration tests are performed in a test facility, and the test structures are examined in terms of dynamic behavior. In the tests, the displacement signal at every grid structure that sustains fuel rods is measured and processed to identify the dynamic properties. The fluid-elastic stability of the structure is also analyzed to evaluate susceptibility to a cross flow with an assumed conservative cross flow distribution. The test and analysis results confirm that the structural integrity can be maintained even in the absence of some welding connections.

IN-PILE PERFORMANCE OF HANA CLADDING TESTED IN HALDEN REACTOR

An in-pile performance test of HANA claddings was conducted at up to 67 GWD/MTU in the Halden research reactor in Norway over a 6.5 year period. Four types of HANA claddings (HANA-3, HANA-4, HANA-5, and HANA-6) and a reference Zircaloy-4 cladding were used for the in-pile test. The evaluation parameters of the HANA claddings were the corrosion behavior, dimensional changes, hydrogen uptake, and tensile strength after the claddings were tested under the simulated operation conditions of a Korean commercial reactor. The oxide thickness ranged from 15 to 37 mm at a high flux region in the test rods, and all HANA claddings showed corrosion resistance superior to the Zircaloy-4 cladding. The creep-down rate of all HANA claddings was lower than that of the Zircaloy-4 cladding. In addition, the hydrogen content of the HANA claddings ranged from 54 to 96 wppm at the high heat flux region of the test rods, whereas the hydrogen content of the Zircaloy-4 cladding was 119 wppm. The tensile strength of the HANA and Zircaloy-4 claddings was similarly increased when compared to the un-irradiated claddings owing to the radiation-induced hardening.

MECHANICAL AND IRRADIATION PROPERTIES OF ZIRCONIUM ALLOYS IRRADIATED IN HANARO

These experimental studies are carried out to build a database for analyzing fuel performance in nuclear power plants. In particular, this study focuses on the mechanical and irradiation properties of three kinds of zirconium alloy (Alloy A, Alloy B and Alloy C) irradiated in the HANARO (High-flux Advanced Neutron Application Reactor), one of the leading multipurpose research reactors in the world. Yield strength and ultimate tensile strength were measured to determine the mechanical properties before and after irradiation, while irradiation growth was measured for the irradiation properties. The samples for irradiation testing are classified by texture. For the irradiation condition, all samples were wrapped into the capsule (07M-13N) and irradiated in the HANARO for about 100 days (E > 1.0 MeV, 1.1 x 10 21 n/cm²). These tests and results indicate that the mechanical properties of zirconium alloys are similar whether unirradiated or irradiated. Alloy B has shown the highest yield strength and tensile strength properties compared to other alloys in irradiated condition. Even though each of the zirconium alloys has a different alloying content, this content does not seem to affect the mechanical properties under an unirradiated condition and low fluence. And all the alloys have shown the tendency to increase in yield strength and ultimate tensile strength. Transverse specimens of each of the zirconium alloys have a slightly lower irradiation growth tendency than longitudinal specimens. However, for clear analysis of texture effects, further testing under higher irradiation conditions is needed.

Vibration Characteristics of a Nuclear Fuel Rod in Uniform Axial Flow

Nuclear fuel rods are exposed to axial flow in a reactor, and flow-induced-vibration due to the flow usually causes damage in the fuel rods. Thus a prior knowledge about dynamic behavior of a fuel rod exposed to the flow condition should be provided. This paper shows that dynamic characteristics of a nuclear fuel rod depend on axial flow velocity. Assuming small lateral displacement, the effects of uniform axial flow are investigated. The analytic results show that axial flow generally reduces fuel rod stiffness and raises its damping in normal condition. Also, the critical axial velocities which make the fuel rod behavior unstable were found. That is, solving generalized eigenvalue equation of the fuel rod dynamic system, the eigenvalues with positive real part are detected. Based on the simulation results, on the other hand, it turns out that the ordinary axial flow in nuclear reactors does not affect to stability of a nuclear fuel rod even in the conservative condition.

Harmonic Response Estimation Method on the Lévy Plate with Two Opposite Edges Having Free Boundary Conditions

This paper discusses a harmonic response estimation method on the Lvy plate with two opposite edges simply supported and the other two edges having free boundary conditions. Since the equation of motion of the plate is not self-adjoint, the modes are not orthogonal to each other on the domain. Noting that the Lvy plate can be expressed using one term sinusoidal function that is orthogonal to other sinusoidal functions, this paper suggested the calculation method that is equivalent to finding a least square error minimization solution of the finite number of algebraic equations. Example problems subjected to a distributed area loading and a distributed line loading are defined and their solutions are provided. The solutions are compared to those of the commercial code, ANSYS. According to the verification results, it is expected that the suggested method will be useful to predict the forced response on the Lvy plate with the distributed area or line loading conditions.

Harmonic Analysis on a Lévy Plate and Its Application to Fatigue Analysis

This paper discusses a harmonic response estimation method on the Levy plate with two opposite edges simply supported and the other two edges having free boundary conditions. Then, the harmonic response is processed to evaluate fatigue damage. Since the equation of motion of the plate is not self-adjoint, the modes are not orthogonal to each other on the structure domain. Noting that the Levy plate can be expressed using one term sinusoidal function that is orthogonal to other sinusoidal functions, this paper suggested the calculation method that is equivalent to finding a least square error minimization solution of the finite number of algebraic equations. Example problems subjected to a distributed area loading and a distributed line loading are defined and their solutions are provided. The solutions are compared to those of the commercial code, ANSYS. The plate motion due to high frequency vibration can be seen in the nuclear fuels subjected to highly turbulent coolant flow. The dominant exciting frequency is dependent on the coolant velocity and Strouhal number, a dimensionless number describing oscillating flow mechanism. This paper also discusses fatigue damage considering the high frequency vibration using the Dirlik equation.

The Grid Strap Vibration Characteristics of the 5×5 Nuclear Fuel Mock-up

Since the fuel is always exposed to turbulent flow, the grid strap shows flow induced vibration characteristics that impact on the nuclear fuel soundness. The dynamic behavior of grids in nuclear fuels is quite complex, since two pairs of spring and dimple support are contacted with rods by friction in the limited space. This paper focuses on investigation of the grid strap(test fuel strap, TFS) vibration in one cell. TFS consists of a single spring and double dimples. To identify the grid strap vibration, modal analysis of the strap is performed using finite element method(FEM). Modal testing on a grid structure without rods is performed. The modal testing results are compared to analytic results. In addition, random test considering rod effect is performed about a grid with rods under real contact condition in the air. Finally, the strap vibration of a fuel bundle in investigation of flow induced vibration(INFINIT) facility is measured in real fluid velocity condition without heating. It is shown that modal frequencies from the test are almost equal to those peak frequencies in the INFINIT test.

CONTACT FORCE MODEL FOR A BEAM WITH DISCRETELY SPACED GAP SUPPORTS AND ITS APPROXIMATED SOLUTION

This paper proposes an approximated contact force model to identify the nonlinear behavior of a fuel rod with gap supports; also, the numerical prediction of interfacial forces in the mechanical contact of fuel rods with gap supports is studied. The Newmark integration method requires the current status of the contact force, but the contact force is not given a priori. Taylor’s expansion can be used to predict the unknown contact force; therefore, it should be guaranteed that the first derivative of the contact force is continuous. This work proposes a continuous and differentiable contact force model with the ability to estimate the current state of the contact force. An approximated convex and differentiable potential function for the contact force is described, and a variational formulation is also provided. A numerical example that considers the particularly stiff supports has been studied, and a fuel rod with hardening supports was also examined for a realistic simulation. An approximated proper solution can be obtained using the results, and abrupt changes from the contacting state to non-contacting state, or vice versa, can be relieved. It can also be seen that not only the external force but also the developed contact force affects the response.

Analysis of Spacer Grid Spring Load-Deflection Characteristic for PWR Fuel Assembly

Spacer grid springs support fuel rods so that the rods keep the position laterally and axially in pressurized water reactor fuel assemblies. The spring load-deflection characteristic, i.e. spring force and stiffness, is needed to evaluate the rod support conditions in the case of fuel assembly manufacturing, shipping and handling. In general, the load-deflection characteristic of grid spring is obtained by mechanical test, but it takes long time to get the new designed grid specimen because the grid manufacturing process comprises strip material manufacturing, stamping die and punch preparation, heat treatment and welding, etc. Therefore the analytic method such as finite element method (FEM) is tried to predict the nonlinear load-deflection characteristic of new designed grid. The spring characteristic mechanical test is simulated with unit cell model and analyzed by FEM tool. Comparing the results between test and analysis shows that more details are needed in the modeling because the boundary conditions of the spring are very complicated and the spring material thickness is changed by the stamping process. The analysis of modified model including expanded cells and thickness changed springs is performed. Using the analytic method of the work to obtain the load-deflection characteristic of spacer grid spring is expected to be useful in the prediction of the characteristic of new designed grids.Copyright