• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.636-643
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• 1995

The accurate determination of the fuel-cladding gap conductance as functions of rod burnup and power level may be a key to the design and safety analysis of a reactor. The incorporation of a sophisticated gap conductance model into nuclear design code for computing thermal hydraulic feedback effect has not been implemented mainly because of computational inefficiency due to complicated behavior of gap conductance. To avoid the time-consuming iteration scheme, simplification of the gap conductance model is done for the current design model. The simplified model considers only the heat conductance contribution to the gap conductance. The simplification is made possible by direct consideration of the gas conductivity depending on the composition of constituent gases in the gap and the fuel-cladding gap size from computer simulation of representative power histories. The simplified gap conductance model is applied to the various fuel power histories and the predicted gap conductances are found to agree well with the results of the design model.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.2488-2493
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• 2007

A gap conductance is very important factor which can affect nuclear fuel temperature. Especially, in case of an annular fuel, a gap conductance effect can lead an unexpected heat split phenomena which is caused by a large difference of an inner and outer gap conductance. The gap conductance mechanism is very complicated behavior due to the its strong dependency on microscopic factors such as a contact surface roughness, local contact pressure and local temperature. In this paper, for the decision of test temperature and pressure range, a procedure and calculation results of in-reactor fuel temperature and pressure analysis are summarized which can be applied to test equipment design and determination of test matrix. Based upon analysis results, it is concluded that the minimum and maximum test temperature are $300^{\circ}C$ and $530^{\circ}C$ respectively, and the maximum pellet/cladding interfacial contact pressure should be observed up to 45MPa.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.2
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• pp.157-166
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• 2014

A light water reactor (LWR) fuel rod consists of zirconium alloy cladding tube and uranium dioxide pellets with a slight gap between them. The modeling of heat transfer across the gap between fuel pellets and the protective cladding is essential to understanding fuel behavior under irradiated conditions. Many researchers have been developing fuel performance codes based on finite element method (FE) to calculate temperature, stress and strain for multidimensional analysis. The gap conductance model for multi-dimension is difficult issue in terms of convergence and nonlinearity because gap conductance is function of gap thickness which depends on mechanical analysis at each iteration step. In this paper, virtual link gap element (VLG) has been proposed to resolve convergence issue and nonlinear characteristic of multidimensional gap conductance. In terms of calculation accuracy and convergence efficiency, the proposed VLG model has been evaluated for variable cases.

• Nuclear Engineering and Technology
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• v.27 no.3
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• pp.344-357
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• 1995

The gap conductance between the fuel and the sheath depends strongly on the gap width and has a significant influence on the amount of initial stored energy. The modified Ross and Stoute gap conductance model in ELESTRES is based on a simplified thermal deformation model for steady-state fuel temperature calculations. A review on a series of experiments reveals that fuel pellets crack relocate, and are eccentrically positioned within the sheath rather than solid concentric cylinders. In this paper, the hue recently-proposed gap conductance models (offset gap model and relocated gap model) are described and are applied to calculate the fuel-sheath gap conductances under experimental conditions and normal operating conditions in CANDU reactors. The good agreement between the experimentally-inferred and calculated gap conductance values demonstrates that the modified Ross and Stoute model was implemented correctly in ELESTRES. The predictions of the modified Ross and Stoute model provide conservative values for gap heat transfer and fuel surface temperature compared to the offset gap and relocated gap models for a limiting power envelope.

• Applied Science and Convergence Technology
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• v.24 no.1
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• pp.9-15
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• 2015

The practical pumping speed of a dry pump is considerably lower than the intrinsic speed because of back-streaming through finite gaps of the rotor assembly. The maximum compression ratio and the ultimate pressure of the pump are also directly influenced by the back-streaming rate. Therefore, information on the gap conductance, which determines the back-streaming characteristics of the rotor assembly, is the most important key for estimating the pumping performance of a dry pump. In this paper, the feasibility of calculating analytically the pumping performance of a multi-stage Roots pump, one of the most popular types of dry pumps, by quantifying the gap conductance in a rational way, is discussed.

• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1629-1635
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• 2017

Reliable prediction of nuclear fuel rod behavior of nuclear power reactors constitutes a basic demand for steady-state calculations, design purposes, and fuel performance assessment. Perfect design of fuel rods as the first barrier against fission product release is very important. Simulation of fuel rod performance with a code or software is one of the fuel rod design steps. In this study, a software program called MARCODE is developed in MATLAB environment that can analyze the temperature distribution, gap conductance value, and fuel and clad displacement in both solid and annular fuel rods. With a comparison of the maximum fuel temperature, fuel average temperature, fuel surface temperature, and gap conductance in solid and annular fuel, the effects of a central hole on the fuel temperature distribution are investigated.

• Progress in Superconductivity
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• v.3 no.2
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• pp.146-150
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• 2002

We performed the point contact spectroscopy on newly discovered superconductor $MgB_2$ thin films with Au tip. In the point contact spectroscopy of the metallic Sharvin limit, the differential conductance below the gap is twice as that above the gap by virtue of Andreev Reflection. After some surface cleaning processes of sample preparation such as ion-milling and wet etching, the obtained dI/dV versus voltage curves are relatively well fitted to the Blonder-Tinkham-Klapwijk (BTK) formalism. Gaps determined by this technique were distributed in the range of 3meV~ 8meV with the BCS value of 5.9meV in the weak coupling limit. We attribute these discrepancies to the symmetry of the gap parameter and the degradation of the surface of the sample. We also present the temperature dependence of the conductance vs voltage curve and thereby the temperature dependence of the gap.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2760-2770
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• 2020

A model for calculation of fuel temperature profile using binary gas mixture of Helium and Xenon for gap gas conductance is proposed here. In this model, the temperature profile of a fuel pencil from fuel centreline to fuel surface has been calculated by taking into account the dilution of Helium gas filled during fuel manufacturing due to accumulation of fission gas Xenon. In this model an explicit calculation of gap gas conductance of binary gas mixture of Helium and Xenon has been carried out. A computer code Fuel Characteristics Calculator (FCCAL) is developed for the model. The phenomena modelled by FCCAL takes into account heat conduction through the fuel pellet, heat transfer from pellet surface to the cladding through the gap gas and heat transfer from cladding to coolant. The binary noble gas mixture model used in FCCAL is an improvement over the parametric model of Lassmann and Pazdera. The results obtained from the code FCCAL is used for fuel temperature calculation in 3-D neutron diffusion solver for the coolant outlet temperature of the core at steady operation at full power. It is found that there is an improvement in calculation time without compromising accuracy with FCCAL.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.3
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• pp.249-257
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• 2015

A fuel assembly consists of fuel rods composed of pellets (UO2) and a cladding tube (Zircaloy). The role of the fuel rods in the reactor is to generate heat by nuclear fission, as well as to retain fission products during operation. A simulation method using a computer program was used to evaluate the safety of the nuclear fuel rods. This computer program has been called the fuel performance code. In the analysis of a light water reactor fuel rod, the gap conductance, which depended on the distance between the pellets and cladding tube, mainly influenced the thermomechanical behavior of the fuel rod. In this work, a 3D gap element was proposed to simulate the thermo-mechanical behavior of the nuclear fuel rod, considering the gap conductance. To implement the proposed 3D gap element, a 3D thermo-mechanical module was also developed using FORTRAN90. The asymmetric characteristics of the nuclear fuel rod, such as the MPS (missing pellet surface) and eccentricity, were simulated to evaluate the proposed 3D gap element.

• Nuclear Engineering and Technology
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• v.23 no.1
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• pp.66-80
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• 1991

The LOFT intermediate break experiment L5-1, which simulates 12 inch diameter ECC line break in a typical PWR, has been analyzed using the reactor thermal/hydraulic analysis code RELAP5/MOD2, Cycle 36.04. The base calculation, which modeled the core with single flow channel and two heat structures without using the options of reflood and gap conductance model, has been successfully completed and compared with experimental data. Sensitivity studies were carried out to investigate the effects of nodalization at reactor vessel and core modeling on major thermal hydraulic parameters, especially on peak cladding temperature(PCT). These sensitivity items are : single flow channel and single heat structure (Case A), two flow channel and two heat structures (Case B), reflood option added (Case C) and both reflood and gap conductance options added (Case D). The code, RELAP5/MOD2 Cycle 36.04 with the base modeling, predicted the key parameters of LOFT IBLOCA Test L5-1 better than Cases A,B,C and D. Thus, it is concluded that the single flow channel modeling for core is better than the two flow channel modeling and two heat structure is also better than single heat structure modeling to predict PCT at the central fuel rods. It is, therefore, recommended to use the reflood option and not to use gap conductance option for this L5-1 type IBLOCA.