• Nuclear Engineering and Technology
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• v.2 no.1
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• pp.19-25
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• 1970

A study was conducted on the coolant mixing between water flowing in two adjacent subchannels. Measurements were made of the quantity of mass transferred between a larger rectangular channel and a smaller triangular channel in a 19-rod fuel bundle under the conditions of single phase flow and air-water two-phase flow. The results of the experiments showed that the low mixing rate appears in single phase flow, and high mixing rate was measured in air-water two-phase flow Mixing rate decreases with the increasing of air void fraction during the air-water flow. It seems that the high mixing rate in the air-water flow was caused due to adequate agitation of the chaotic air void.

• Nuclear Engineering and Technology
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• v.34 no.6
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• pp.596-609
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• 2002

Korea Electric Power Research Institute has developed the in-house safety analysis methodologies for non-LOCA(Loss Of Coolant Accident) events based on codes and methodologies of vendors and Electric Power Research Institute . According to the new methodologies, analyses of system responses and calculation of DNBR(Departure from Nucleate Boiling Ratio) during the transient have been carried out with RETRAN code and a sub-channel analysis code, respectively. However, it takes too much time to calculate DNBR for each case using the two codes to search for the limiting case from sensitivity study. To simplify the search for the limiting case, accordingly, RETRAN code has been modified to roughly calculate DNBR using hot channel modeling. The W-3 correlation is already included in RETRAN as one of the auxiliary DNBR models. However, WRB-1 and WRB-2 correlations required to analyze some Westinghouse type fuels are not considered in RETRAN DNBR models. In this paper, the RETRAN DNBR models using the correlations have been developed and the partial and complete loss of forced reactor coolant flow events have been analyzed for Yonggwang units 1 and 2 with the new methodologies to validate the models. The results of the analyses have been compared with those mentioned in the chapter 15 of the Final Safety Analysis Report.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.589-596
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• 2013

In a pressurized heavy water reactor, following loss of the primary coolant, severe core damage would begin with the depletion of the liquid moderator, exposing the top row of internally-voided fuel channels to steam cooling conditions on the inside and outside. The uncovered fuel channels would heat up, deform and disassemble into core debris. Large inventories of water passively reduce the rate of progression of the accident, prolonging the time for complete loss of engineered heat sinks. The efficacy of available backup and ultimate heat sinks, available in a CANDU 6 reactor, in mitigating the consequences of a prolonged station blackout scenario was analysed using the MAAP4-CANDU code. The analysis indicated that the steam generator secondary side water inventory is the most effective heat sink during the accident. Additional heat sinks such as the primary coolant, moderator, calandria vault water and end shield water are also able to remove decay heat; however, a gradually increasing mismatch between heat generation and heat removal occurs over the course of the postulated event. This mismatch is equivalent to an additional water inventory estimated to be 350,000 kg at the time of calandria vessel failure. In the Enhanced CANDU 6 reactor ~2,040,000 kg of water in the reserve water tank is available for prolonged emergencies requiring heat sinks.

• Journal of the Korean Society of Propulsion Engineers
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• v.24 no.5
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• pp.1-12
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• 2020

In this paper, the heat transfer characteristics of Jet A-1, which is used as a coolant and fuel in a regeneratively cooled thrust chamber, were experimentally studied. By varying the applied current for heating the cooling channel, the simulated specimen diameter, the specimen outlet pressure and the coolant flow rate, the wall temperatures of the specimen and the Jet A-1 temperatures at the specimen inlet/outlet were measured. It was found that the specimen diameter and the flow rate were important factors for the characteristics of heat transfer and the outlet pressure did not affect the performance of heat transfer. The results of the heat transfer experiments were compared with the previous Nusselt number empirical equations and novel Nusselt number empirical equations were finally derived.

• Journal of the Korean Society of Manufacturing Process Engineers
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• v.21 no.6
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• pp.44-49
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• 2022

The improvement in the battery performance and life using a battery thermal management system directly affects the improvement in the performance, life, and energy efficiency of electric vehicles. Therefore, this study numerically analyzed the heat exchange processes between the coolant inside the cooling plate channel and the heat generated by the battery. The cooling performance was analyzed based on the average temperature, temperature uniformity, and the maximum and minimum temperature differences of the battery. A performance difference existed depending on the coolant inlet temperature but showed the same tendency of cooling performance according to the shape of each plate's channel. Type 1 showed the best results in terms of battery temperature uniformity, which is the most important measure of battery performance; Type 2 showed the best results in terms of the average temperature of the battery; and Type 3 showed the best results in terms of the maximum and minimum temperature differences of the battery compared with that of the other cooling plates.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.35-40
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• 1995

Considering the higher discharge burnup, lower channel refuelling rate, lower linear element rating(LER), lower coolant void reactivity and axial power shape, CANFLEX-RU fuel bundle is optimized for CANDU-6 by grading the fissile composition in the ring-wise of the bundle and by applying fuel management scheme appropriately. The fissile composition of the fuel bundle is graded as the recovered uranium (0.9 w/o U-235) in the outer and intermediate elements, depleted Uranium (0.2 w/o U-235) in the center element, natural uranium (0.71 w/o U-235) in the inner elements. Enrichment is not required for these fuel. The fissile composition is optimized by lattice calculation and by time-averaged reactor simulation. CANFLEX-RU optimized for CANDU-6 resulted to be the 15% lower channel refuelling rate, acceptable axial power profile and power envelope, 70% higher discharge burnup, 15% lower LER and not increase coolant void reactivity compared with the 37-element natural uranium bundle for CANDU-6.

• 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 Korean Society for Noise and Vibration Engineering Conference
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• 2003.11a
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• pp.345-349
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• 2003

The dynamic behavior of CANDU nuclear fuel channel was analyzed by the use of 3-dimensional finite element method, under the various fault conditions such as a fault in the end fitting support and the removal/migration of the garter spring in the fuel channel, in order to predict the dynamic behavior for a degraded symptoms of CANDU nuclear fuel channel. Moreover, the frequency response analysis for possible fault conditions was also peformed considering the effects of the pressure tube vibration and flow-induced vibration by the coolant flow. From the analysis of the frequency responses, defects in the garter spring have influenced the changes of 2nd and 3rd modes and all the important modes are varied for the failure in the journal bearing in the end fitting body.

• Journal of computational fluids engineering
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• v.12 no.4
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• pp.74-84
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• 2007

The flowfield characteristics in a straight rectangular channel have been investigated through a numerical model to analyze the regenerative cooling system that is used in rocket engine cooling. The supercritical hydrogen coolant introduces strong property variations that have a major influence on the developing flow and heat transfer characteristics. Of particular interest is the improved understanding of the physical characteristics of such flows through parametric studies. The approach used is a numerical solution of the full Navier-Stokes equations in the three dimensional form including the arbitrary equation of state and property variations. The present study compares constant and variable property solutions for both laminar and turbulent flow. For laminar flow, the variation of aspect ratio is examined, while for turbulent flow, the effects of variation of channel length and Reynolds number are discussed.

• Journal of the Korean Society for Precision Engineering
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• v.28 no.1
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• pp.63-72
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• 2011

In recent years, hot-stamped components are more increasingly used in the automotive industry in order to reduce weight and to improve the strength of vehicles. In hot stamping process, blank is hot formed and press hardened in a tool. However, in hot stamping without cooling channel, temperature of the tool increases gradually in mass production thus cannot meet the critical cooling rate to obtain high strength over 1500MPa. Warpage occurs in the hot stamped component due to non-uniform stress state caused by unbalanced cooling. Therefore, tools should be uniformly as well as rapidly cooled down by the coolant which flows through cooling channel. In this paper, optimal design method of cooling channel to obtain uniform and high strength of the component is proposed. Optimized cooling channel is applied to the hot press V-bending process. As a result of measuring strength, hardness and microstructure of the hot formed parts, it is known that the design methodology of cooling channel is effective to the hot stamping process.