• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3207-3216
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• 2021

Three-dimensional structures of a vortical flow field and heat transfer characteristics in a partially blocked 7-pin fuel assembly mock-up of sodium-cooled fast reactor have been investigated through a numerical analysis using a commercial computational fluid dynamics code, ANSYS CFX. The simulation with the SST turbulence model agrees well with the experimental data of outlet and cladding wall temperatures. From the analysis on the limiting streamline at the wall, multi-scale vortexes developed in axial direction were found around the blockage. The vortex core has a high cladding wall temperature, and the attachment line has a low cladding wall temperature. The small-scale vortex structures significantly enhance the convective heat transfer because it increases the turbulent mixing and the turbulence kinetic energy. The large-scale vortex structures supply thermal energy near the heated cladding wall surface. It is expected that control of the vortex structures in the fuel assembly plays a significant role in the convective heat transfer enhancement. Furthermore, the blockage plate and grid spacer increase the pressure drop to about 36% compared to the bare case.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1487-1503
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• 2019

The methods and performance of a pin-level nuclear reactor core thermal-hydraulics (T/H) code ESCOT employing the drift-flux model are presented. This code aims at providing an accurate yet fast core thermal-hydraulics solution capability to high-fidelity multiphysics core analysis systems targeting massively parallel computing platforms. The four equation drift-flux model is adopted for two-phase calculations, and numerical solutions are obtained by applying the Finite Volume Method (FVM) and the Semi-Implicit Method for Pressure-Linked Equation (SIMPLE)-like algorithm in a staggered grid system. Constitutive models involving turbulent mixing, pressure drop, and vapor generation are employed to simulate key phenomena in subchannel-scale analyses. ESCOT is parallelized by a domain decomposition scheme that involves both radial and axial decomposition to enable highly parallelized execution. The ESCOT solutions are validated through the applications to various experiments which include CNEN $4{\times}4$, Weiss et al. two assemblies, PNNL $2{\times}6$, RPI $2{\times}2$ air-water, and PSBT covering single/two-phase and unheated/heated conditions. The parameters of interest for validation include various flow characteristics such as turbulent mixing, spacer grid pressure drop, cross-flow, reverse flow, buoyancy effect, void drift, and bubble generation. For all the validation tests, ESCOT shows good agreements with measured data in the extent comparable to those of other subchannel-scale codes: COBRA-TF, MATRA and/or CUPID. The execution performance is examined with a mini-sized whole core consisting of 89 fuel assemblies and for an OPR1000 core. It turns out that it is about 1.5 times faster than a subchannel code based on the two-fluid three field model and the axial domain decomposition scheme works as well as the radial one yielding a steady-state solution for the OPR1000 core within 30 s with 104 processors.

• Journal of Mechanical Science and Technology
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• v.14 no.8
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• pp.816-822
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• 2000

In the core of the nuclear power plant of PWR, several cases of fuel failure by unknown causes have been experienced for various fuel types. From the common features of the failure pattern, failure lead time, flow conditions, and flow induced vibration characteristics in nuclear fuel bundles, it is deduced that the fretting wear failure of the fuel rod at the spacer grid position is due to the fluidelastic vibration. In the past, fluidelastic vibration was simulated by quasi -static semi-analytical model, so called the static model, which could not account for the interaction between the rods within a bundle. To overcome this defect and to provide for more flexibilities applicable to the fuel bundle, Tanaka's unsteady model was modified to accomodate the geometrical differences and governing parameter changes during the operations such as the number of rods, pitch to diameter ratio (P/D), spring force, damping coefficient, etc. The critical velocity was calculated by solving the governing equations with the MATLAB code. A comparison between the estimated critical velocity and the test result shows a good agreement. Finally, the level of decrease of the critical velocity due to the reduction in the spring force and reduced damping coefficient due to the radiation exposure is also estimated.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.28 no.5
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• pp.186-194
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• 2016

As a turbulence-enhancing device, a mixing vane installed at a spacer grid of the fuel assembly plays a role in improving the convective heat transfer by generating either swirl flow in the subchannels or cross flow between fuel rod gaps. Therefore, both configuration and arrangement pattern of a mixing vane are important factors that determine the performance of a mixing vane. In this study, in order to examine the flow distribution features inside $5{\times}5$ fuel assembly with swirl-type mixing vanes used in benchmark calculation of OECD/NEA, simulations were conducted with commercial CFD software ANSYS CFX R.14. Predicted results were compared to data measured from MATiS-H (Measurement and Analysis of Turbulent Mixing in Subchannels-Horizontal) test facility. In addition, the effect of swirl-type mixing vanes on flow pattern inside the fuel assembly was described.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.590-595
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• 2003

An Exponential experiment system which is composed of a neutron detector, a signal analysis system and a neutron source, Cf-252 has been installed in order to experimentally determine the neutron effective multiplication factor for a PWR spent fuel assembly. The axial background neutron flux is measured in a preliminary performance test. From the results, the spacer grid position is determined to be consistent with the design specifications within a 2.3% relative error. The induced fission neutron for four of the assemblies is also measured by scanning the neutron source, Cf-252 or the neutron detector. The exponential decay constants have been evaluated by the application of the Poisson regression to the net induced fission neutron counts. The measured keffs determined on the basis of the exponential decay constants of Cl5 appeared to be 0.541, 0.540, 0.597 and 0.556, respectively, which are comparable with 0.55195$\pm$0.00232 of the MCNP calculation.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.534-539
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• 1997

A CHF table method has been assessed in this study for rod bundle CHF predictions. At the conceptual design stage for a new reactor, a general critical heat flux (CHF) prediction method with a wide applicable range and reasonable accuracy is essential to the thermal-hydraulic design and safety analysis. In many aspects, a CHF table method (i.e., the use of a round tube CHF table with appropriate bundle correction factors) can be a promising way to fulfill this need. So the assessment of the CHF table method has been performed with the bundle CHF data relevant to pressurized water reactors (PWRs). For comparison purposes, W-3R and EPRI-1 were also applied to the same data base. Data analysis has been conducted with the subchannel code COBRA-IV-I. The CHF table method shows the best predictions based on the direct substitution method. Improvements of the bundle correction factors, especially for the spacer grid and cold wall effects, are desirable for better predictions. Though the present assessment is somewhat limited in both fuel geometries and operating conditions, the CHF table method clearly shows potential to be a general CHF predictor.

• Nuclear Engineering and Technology
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• v.33 no.6
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• pp.619-625
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• 2001

An axial-flow-induced vibration model was proposed for a rod supported by two translational springs at both ends. For developing the model, a one-mode approximation was made based on the assumption that the first mode was dominant in vibration behavior of the single span rod. The first natural frequency and mode shape functions for the flow-induced vibration, called the FIV model were derived by using Lagrange＇s method. The vibration displacements at reactor conditions were calculated by the proposed model for the spring-supported rod and by the previous model for the simple-supported(55) rod. As a result, the vibration displacement for the spring-supported rod was larger than that of the 55 rod, and the discrepancy between both displacements became much larger as flow velocity increased. The vibration displacement for the spring-supported rod appeared to decrease with the increase of the spring constant. AS flow velocity increased, the increase rate of vibration displacement was calculated to go linearly up, and that of the rod having the short span length was larger than that of the rod having the long span length although the displacement value itself of the long span rod was larger than that of the short one.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.1079-1092
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• 2020

A parametric study of several parameters relevant to design safety on the spent nuclear fuel (SNF) rod response under a drop accident is presented. In the view of the complexity of interactions between the independent safety-related parameters, a factorial design of experiment is employed as an efficient method to investigate the main effects and the interactions between them. A detailed single full-length fuel rod is used with consideration of post-irradiated fuel conditions under horizontal and vertical free-drops onto an unyielding surface using finite-element analysis. Critical drop heights and critical g-loads that yield the threshold plastic strain in the cladding are numerically estimated to evaluate the fuel rod structural resistance to impact load. The combinatory effects of four uncertain parameters (pellet-cladding interfacial bonding, material properties, spacer grid stiffness, rod internal pressure) and the interactions between them on the fuel rod response are investigated. The principal finding of this research showed that the effects of above-mentioned parameters on the load-carrying capacity of fuel rod are significantly different. This study could help to prioritize the importance of data in managing and studying the structural integrity of the SNF.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1218-1230
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• 2019

A sub-channel solver, named ${\underline{S}}teady$ and ${\underline{T}}ransient$ ${\underline{A}}nalyzer$ for ${\underline{R}}eactor$ ${\underline{T}}hermal$ hydraulics (START), has been developed using the homogenous model for two-phase conditions of light water reactors. The code is developed as a fast and accurate TH-solver for coupled and multi-physics calculations. START has been validated against the NUPEC PWR Sub-channel and Bundle Test (PSBT) database. Tests like single-channel quality and void-fraction for steady state, outlet fluid temperature for steady state, rod-bundle quality and void-fraction for both steady state and transient conditions have been analyzed and compared with experimental values. Results reveal a good accuracy of solution for both steady state and transient scenarios. Axially different values for turbulent mixing coefficient are used based on different grid-spacer types. This provides better results as compared to using a single value of turbulent mixing coefficient. Code-to-code evaluation of PSBT results by the START code compares well with other industrial codes. The START code has been parallelized with the OpenMP algorithm and its numerical performance is evaluated with a large whole PWR core. Scaling study of START shows a good parallel performance.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.35 no.8
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• pp.889-897
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• 2011

In this work, we investigated the theoretical forming limit models for Zircaloy-4 and Zirlo used for spacer grid of nuclear fuel rods. Tensile and anisotropy tests were performed to obtain stress-strain curves and anisotropic coefficients. The experimental forming limit diagrams (FLD) for two materials were obtained by dome stretching tests following NUMISHEET 96. Theoretical FLD depends on FL models and yield criteria. To obtain the right hand side (RHS) of FLD, we applied the FL models (Swift's diffuse necking, M-K theory, S-R vertex theory) to Zircaloy-4 and Zirlo sheets. Hill's local necking theory was adopted for the left hand side (LHS) of FLD. To consider the anisotropy of sheets, the yield criteria of Hill and Hosford were applied. Comparing the predicted curves with the experimental data, we found that the RHS of FLD for Zircaloy-4 can be described by the Swift model (with the Hill's criterion), while the LHS of the FLD can be explained by Hill model. The FLD for Zirlo can be explained by the S-R model and the Hosford's criterion (a = 8).