• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.237-246
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• 1994

A high flow critical heat flux (CHF) correlation, based on the single-pin CHF experimental data for finned and unfinned heated rods, was developed for the thermal-hydraulic design and safety analysis of the Korea Multi-purpose Research Reactor (KMRR) core. The correlation consists of dimensionless parameters such as Reynolds number, thermodynamic equilibrium quality, liquid-to-vapor density ratio, and hydraulic equivalent diameter ratio. The fin effect was taken into account in the correlation by a finned-to-unfinned heated perimeter ratio. The effects of a cold wall and non-uniform axial power distribution ore discussed to verify the applicability of the single-pin based correlation to the KMRR fuel bundle. The correlation limit departure from nucleate boiling ratio (DNBR) was determined as 1.44 from the statistical analysis of the CHF data.

• Journal of Energy Engineering
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• v.7 no.1
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• pp.17-23
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• 1998

A simplified experiment was performed to figure out the atomization characteristics of highly viscous liquid of rice-bran oil by applying ultrasonic energy to improve the atomization of spray droplets. A spray system, an ultrasonic system, and three kinds of pintle-type nozzles(pin-edge angle: 5 , 10 , 15 ) were manufactured. To investigate the effects of ultrasonic energy on the atomization of a highly viscous liquid, a phase doppler particle analyzer was used for the measurement and calculation of spray droplets data. Nozzle opening pressures were chosen of 3 levels, i.e, 10, 13, 16 MPa. As a result, it could be concluded that the ultrasonic energy was effective to improve the spray atomization when applied to the fuel by means of 3 different nozzles because of the effects of the liquid fuel cavitation and relaxation between molecules caused by ultrasonic energy. The improvement rate of the spray atomization by the ultrasonic spray atomization by the ultrasonic spray compared with the conventional spray was about 10% increase in the case of pintle type nozzles. With the increase of pin-edge angles the distribution lines by nozzle opening pressures are declined for both conventional and ultrasonic sprays. This means that the increase of the pin-edge angle improves the atomization of sprays.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.677-690
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• 2009

A previous study demonstrated that the two-row fuel assembly has much more favorable neutron-physical and thermal-hydraulic behavior than the conventional one-row fuel assemblies. Based on the newly developed two-row fuel assembly, an SCWR core is proposed and analyzed. The performance of the proposed core is investigated with 3-D coupled neutron-physical and thermal-hydraulic calculations. During the coupling procedure, the thermal-hydraulic behavior is analyzed using a sub-channel analysis code and the neutron-physical performance is computed with a 3-D diffusion code. This paper presents the main results achieved thus far related to the distribution of some neutronic and thermal-hydraulic parameters. It shows that with adjustment of the coolant and moderator mass flow in different assemblies, promising neutron-physical and thermal-hydraulic behavior of the SCWR core is achieved. A sensitivity study of the heat transfer correlation is also performed. Since the pin power in fuel assemblies can be non-uniform, a sub-channel analysis is necessary in order to investigate the detailed distribution of thermal-hydraulic parameters in the hottest fuel assembly. The sub-channel analysis is performed based on the bundle averaged parameters obtained with the core analysis. With the sub-channel analysis approach, more precise evaluation of the hot channel factor and maximum cladding surface temperature can be achieved. The difference in the results obtained with both the sub-channel analysis and the fuel assembly homogenized method confirms the importance of the sub-channel analysis.

• Proceedings of the KSME Conference
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• 2004.11a
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• pp.383-387
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• 2004

The Fuel Test Loop (FTL) consists of In-Pile Test Section (IPS) and Out-of-Pile System (OPS). The test condition in IPS such as pressure, temperature and quality of the main cooling water, can be controlled by the OPS. The FTL has been developed to be able to irradiate three pins to the core irradiation hole (IR1 hole) by considering for its utility and user's irradiation requirement. The IPS vessel assembly (IVA) consists of IPS head, outer pressure vessel, inner pressure vessel, inner assembly and test fuel carrier. The IVA is approximately 5.6 m long and fits within a 74 mm in diameter envelope over the full height of the chimney. Above the top of the chimney, the head of the IPS is enlarged to allow the closure flanges and pipe work connections. IVA was designed to test the CANDU and PWR nuclear fuel pin together. Specially, wished to minimize interference by nuclear fuel change in design and synthesize these items and shape design for IVA.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2000.11a
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• pp.441-444
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• 2000

The recycling process for reuse of uranium in the spent fuels consists various unit processes and the decladding process to extract the spent fuel pellet from the zirconium-based cladding is the beginning process of the recycling. There are two methods - mechanical and chemical - in the decladding process. In this paper, the mechanical decladding device by using a motor as a driving part and a press pin to separate the pellets from tube has been developed. This device was automated and modularized to make the remote operation and maintenance easy.

• Journal of Mechanical Science and Technology
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• v.15 no.3
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• pp.403-409
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• 2001

The main objective of the present study is to perform an experimental evaluation of five existing correlations for the subchannel pressure drop analysis of a wire-wrapped fuel assembly. For this purpose, a series of water experiments have been performed using a helical wire-wrapped 19-pin fuel assembly for various test parameters. Four different test sections with different pitch to rod diameter ratios (P/D) and wire lead length to rod diameter ratios (H/D) have been fabricated. A series of pressure drop measurements were made to obtain friction factors for these four test sections. The new data along with existing data are used to evaluate existing correlations. Both the original and the simplified Cheng and Todreas correlations give the best agreement with experimental data for all flow regions.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2006.05a
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• pp.250-253
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• 2006

The in-pile Section (IPS) is subjected to flow-induced vibration(FIV) due to the flow of the primary coolant and then the structural integrity. The in-pile Section (IPS) of 3-pin Fuel Test Loop(FTL) shall be installed in the vortical hole call IR1 of HANARO reactor core. In order to verify the velocity and displacement both the inside region of IPS at the annular region of IPS, the vibration was measured by varing the flow rate on both regions. The displacements of fuel assembly in the in-pile Section (IPS) were found to be lower than the values of allowable design criteria.

• Tribology and Lubricants
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• v.31 no.5
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• pp.189-194
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• 2015

This paper focuses on the conformal contact problem. A typical example of conformal contact is the contact between a pin and hole. In particular, this paper focuses on the condition for assuming a contact stress field to be a Hertzian pressure profile by using well-known classical solutions associated with Hertzian contact. Persson first developed the conformal contact analysis method around half a century ago, but there have been no significant improvements since then. The present research also adopted this method, but developed new solutions from the viewpoint of application to structural design. The analysis began with a comparison between Persson°Øs conformal contact stress and the Hertzian stress fields. The next step was to check the differences in the normalized stress values of both. This study used the tolerance for the difference in the peak stresses of Persson°Øs solution and the Hertz solution to validate the Hertzian assumption. This gave the range for the difference in radii of the pin and hole when the contact force and mechanical properties of the material are specified. The results showed that, at a tolerance of 5%, the Hertzian assumption is valid if half of the contact angle is less than 35°ý. In addition, the Hertzian assumption holds even for a relatively long contact length, in contrast to the general incomplete contact problem. This paper discusses these results along with other aspects of the application to the design.

• Nuclear Engineering and Technology
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• v.41 no.2
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• pp.171-178
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• 2009

The rate of delayed hydride cracking (DHC), V, has been measured in cold-worked and stress-relieved Zircaloy-4 fuel cladding using the Pin-Loading Tension technique. At $250^{\circ}C$ the mean value of V from 69 specimens was $3.3({\pm}0.8)x10^{-8}$ m/s while the temperature dependence up to $275^{\circ}C$ was described by Aexp(-Q/RT), where Q is 48.3 kJ/mol. No cracking or cracking at very low rates was observed at higher temperatures. The fracture surface consisted of flat fracture with no striations. The results are compared with previous results on fuel cladding and pressure tubes.

• Journal of Energy Engineering
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• v.13 no.3
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• pp.181-190
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• 2004

Computational models for analyzing the in-reactor behavior of metallic fuel pins under transient conditions in liquid-metal reactors are developed and implemented in the TRAMAC (TRAnsient thermo-Mechanical Analysis Code) for a metal fuel rod under transient operation conditions. Not only the basic models for a fuel rod performance but also some sub-models used for transient condition are installed in TRAMAC. Among the models, a fission gas release model, which takes the multi-bubble size distribution into account to characterize the lenticular bubble shape and the saturation condition on the grain boundary and the cladding deformation model have been developed based mainly on the existing models in the MAC-SIS code. Finally, cladding strains are calculated from the amount of thermal creep, irradiation creep, and irradiation swelling. The cladding strain model in TRAMAC predicts well the absolute magnitudes and gen-eral trends of their predictions compared with those of experimental data. TRAMAC results for the FH-1,2,6 pins are more conservative than experimental data and relatively reasonable than those of FPIN2 code. From the calculation results of TRAMAC, it is apparent that the code is capable of predicting fission gas release, and cladding deformation for LMR metal fuel finder transient operation conditions. The results show that in general, the predictions of TRAMAC agree well with the available irradiation data.