• Nuclear Engineering and Technology
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• v.23 no.4
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• pp.401-409
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• 1991

This work is concerned with the development of an analytical model to predict the friction in longitudinally finned rod bundles. Such bundles are currently considered in KMRR design. The present model assumes the validity of the Law of the Wall over entire flow area. The flow channel area is divided into the interfin region and a number of element channels, and the algebraic form of the Law of the Wall is integrated over each element channel and interfin region to yield an analytic expression for the pressure drop. The model reasonably predicts the 6 fin KMRR data, and overpredicts the 8 fin data about 15 percent.

• Nuclear Engineering and Technology
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• v.41 no.7
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• pp.907-920
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• 2009

This study was performed to evaluate the prediction capability of a commercial CFD code and to investigate the effects of different geometries such as a 4.4 mm tube and an 8/10 mm annular channel on the detailed flow structures. A numerical simulation was performed for the conditions, at which the experimental data was produced by the test facility SPHINX. A 2-dimensional axisymmetric steady flow was assumed for computational simplicity. The RNG $\kappa-\varepsilon$ turbulence model (RNG) with an enhanced wall treatment option, SST $\kappa-\omega$ (SST) and low Reynolds Abid turbulence model (ABD) were employed and the numerical predictions were compared with the experimental data generated from the experiment. The effects of the geometry on heat transfer were investigated. The flow and temperature fields were also examined in order to investigate the mechanism of heat transfer near the wall. The local heat transfer coefficient predicted by the RNG model is very close to the measurement result for the tube. In contrast, the local heat transfer coefficient predicted by the SST and ABD models is closer to the measurement for the annular channel.

• Nuclear Engineering and Technology
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• v.46 no.2
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• pp.195-206
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• 2014

Heat transfer characteristics in a narrow rectangular channel are experimentally investigated for upward and downward flows. The experimental data obtained are compared with existing data and predictions by many correlations. Based on the observations, there are differences from others: (1) there are no different heat transfer characteristics between upward and downward flows, (2) most of the existing correlations under-estimate heat transfer characteristics, and (3) existing correlations do not predict the high heat transfer in the entrance region for a wide range of Re. In addition, there are a few heat transfer correlations applicable to narrow rectangular channels. Therefore, a new set of correlations is proposed with and without consideration of the entrance region. Without consideration of the entrance region, heat transfer characteristics are expressed as a function of Re and Pr for turbulent flows, and as a function of Gz for laminar flows. The correlation proposed for turbulent and laminar flows has errors of ${\pm}18.25$ and ${\pm}13.62%$, respectively. With consideration of the entrance region, the heat transfer characteristics are expressed as a function of Re, Pr, and $z^*$ for both laminar and turbulent flows. The correlation for turbulent and laminar flows has errors of ${\pm}19.5$ and ${\pm}22.0%$, respectively.

• Earthquakes and Structures
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• v.8 no.3
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• pp.779-799
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• 2015

The slosh height and the possibility of water spill from rectangular Spent Fuel Storage Bays (SFSB) and Tray Loading Bays (TLB) of Nuclear power plant (NPP) are studied during 0.2 g, Safe Shutdown Earthquake (SSE) level of earthquake. The slosh height obtained through Computational Fluid dynamics (CFD) is compared the values given by TID-7024 (Housner 1963) and American concrete institute (ACI) seismic codes. An equivalent amplitude method is used to compute the slosh height through CFD. Numerically computed slosh height for first mode of vibration is found to be in agreement the codal values. The combined effect in longitudinal and lateral directions are studied separately, and found that the slosh height is increased by 24.3% and 38.9% along length and width directions respectively. There is no liquid spillage under SSE level of earthquake data in SFSB and TLB at convective level and at free surface acceleration data. Since seismic design codes do not have guidelines for combined excitations and effect of higher modes for irregular geometries, this CFD procedure can be opted for any geometries to study effect of higher modes and combined three directional excitations.

• Nuclear Engineering and Technology
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• v.28 no.3
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• pp.265-273
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• 1996

Among those tests performed during the Yonggwang Nuclear Power Plant Units 3 and 4 (YGN 3&4) Power Ascension Test period, the Loss of a Main Feedwater Pump test at l00% power is one of the major test which characterize the capability of YGN 3&4. In this event, one of the two normally operating main feedwater pumps is tripped resulting in a 50% reduction in the feedwater flow. Unless the NSSS and Turbine/Generator control systems actuate properly, the reactor will be tripped on low SG water level or high pressurizer pressure. The test performed at Unit 3 was successful by meeting all acceptance criteria, and the plant was stabilized at a reduced power level without reactor trip. The measured test data for the major plant parameters are compared with the predictions made by the KISPAC computer code, an updated best-estimate plant performance analysis code, to verify and validate its applicability. The comparison results showed good agreement in the magnitude as well as the trends of the major plant parameters. Therefore, the KISPAC code can be utilized for the best-estimate nuclear power plant design and simulation tool after a further verification using other plant test data.

• Nuclear Engineering and Technology
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• v.29 no.1
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• pp.35-44
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• 1997

Critical too-Phase flow rates of subcooled water through Short Pipes (L 140039n) with small diameters (D$\leq$7.15 min) have been experimentally investigated for wide ranges of subcooling (0~199$^{\circ}C$) and pressure (0.5~2.0 MPa). To examine the effects of various parameters (i.e., the location of flashing inception, the degree of subcooling, the stagnation temperature and pressure, and the pipe size) on the critical two-phase flow rates of subcooled water through short pipes with small diameters, a total of 135 runs were made for various combinations of test parameters using four different L/D test sections. Experimental results that show effect of various parameters on subcooled critical two phase flow rates are presented in the form of graphs such as the dimensionless mass flux ( $G^{*}$) versus the dimensionless subcooling ( $T_{sub}$$^{*}$) curve. An empirical correlation expressed in terms of a dimensionless subcooling is also obtained for subcooled two-phase flow rates through present test sections. Comparisons between the mass fluxes calculated by present correlation and a total of 755 selected experimental data points of 9 different investigators show that the agreement is fairly good except for very low subcooling data obtained from small L/D (less than 10) orifices.s.s.s.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.247-247
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• 1999

This paper describes a new mechanistic thermal conductivity model considering the heterogeneous microstructure of MOX fuel. Even though the thermal conductivities of MOX have been investigated numerously by experimental measurements and theoretical analyses, they show the large scattering making the performance analysis of MOX fuel difficult. Therefore, a thermal conductivity model that depends on the heterogeneous microstructure of MOX fuel has been developed by using a general two-phase thermal conductivity model. In order to apply this model for developing the thermal conductivity for heterogeneous MOX fuel, the fuel is assumed to consist of Purich particles and U02 matrix including Pu02 in solid solution. Since little relevant data on Purich particles is available, FIGARO and SiemensKWU results are only used to characterize the microstructure of unirradiated and irradiated fuel. Philliponneaus and HALDEN models are selected for the local thermal conductivities for Purich particles and matrix, respectively. Then by combining the two models, overall thermal conductivity of MOX fuel is obtained. The new proposed model estimates the MOX thermal conductivity about 10% less than the value of U02 fuel, which is in the range of MOX thermal conductivity from HALDEN. The developed thermal conductivity model has been incorporated into KAERIs fuel performance code, COSMOS, and then verified using the measured data in the FIGARO program. Comparison of predicted and measured temperatures shows the reasonable agreement within acceptable error bounds together with satisfactory results for the fission gas release and gap pressure.essure.

• Nuclear Engineering and Technology
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• v.35 no.2
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• pp.108-120
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• 2003

In assessing the long term post closure radiological safety assessment of a potential HLW repository in Korea, three categories of uncertainties exist. The first one is the scenario uncertainty where series of different natural events are translated into written statements. The second one is the modeling uncertatinty where different mathematical models are applied for an identical scenario. The last one is the data uncertainty which can be expressed in terms of probabilistic density functions. In this analysis, three different scenarios are seleceted; a small well scenario, a radiolysis scenario, and a naturally discharged scenario. The MASCOT-K and the AMBER, probabilistic safety assessment codes based on connection of sub-modules and a compartment theory respectively, are applied to assess annual individual doses for a generic biosphere. Results illustrate that for a given scenario, predictions from two different codes fairly match well each other But the discrepancies for the different scenarios are significant. However, total doses are still well below the guideline of 2 mRem/yr. Detailed analyses with model and data uncertainties are underway to further assure the safety of a Korean reference dispsoal concept.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.248.1-248
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• 1999

Preliminary analysis of in-reactor thermal performance of three MOX fuel rods, which are going to be irradiated in the Halden reactor beginning in the first Quarter of the year 2000 under the framework of the OECD Halden Reactor Programme, have been conducted by using the computer code COSMOS to ensure their safe operation. Parametric studies have been carried out to investigate the effect of uncertainties on in-reactor behavior by considering the four kinds of uncertainties; thermal conductivity, linear power, manufacturing parameters, and model constants. The analysis shows that, in the case of annular MOX -1 fuel, calculation results for thermal performance vary widely depending on the selection of model constants for fission gas release (FGR). On the contrary, the thermal performance of solid MOX - 3 fuel does not depend on the choice of FGR constants to a large extent as MOX-I, because the fuel temperature is very high in the MOX-3 irrespective of the choice of FGR constants and hence the capacity of grain boundaries to retain gas atoms is not large enough to accommodate the number of gas atoms reaching the grain boundaries. It is planned that when the data on microstructure and thermal conductivity for each type of MOX fuel are available, new analysis will be made using these information. In addition, FGR model constants will be derived from the measured fuel centerline temperature, rod internal pressure and other related data.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4022-4032
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• 2021

In this study, machine learning (ML) techniques were used to model surveillance test data of nuclear power plants from an international database of the ASTM E10.02 committee. Regression modeling was conducted using various techniques, including Cubist, XGBoost, and a support vector machine. The root mean square deviation of each ML model for the baseline dataset was less than that of the ASTM E900-15 nonlinear regression model. With respect to the interpolation, the ML methods provided excellent predictions with relatively few computations when applied to the given data range. The effect of the explanatory variables on the transition temperature shift (TTS) for the ML methods was analyzed, and the trends were slightly different from those for the ASTM E900-15 model. ML methods showed some weakness in the extrapolation of the fluence in comparison to the ASTM E900-15, while the Cubist method achieved an extrapolation to a certain extent. To achieve a more reliable prediction of the TTS, it was confirmed that advanced techniques should be considered for extrapolation when applying ML modeling.