• Nuclear Engineering and Technology
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• 제44권4호
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• pp.355-362
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• 2012

The aim of this study was to develop a bundle position-wise linear model (BPLM) to predict Pressure Tube (PT) diametral creep employing the previously measured PT diameters and operating conditions. There are twelve bundles in a fuel channel, and for each bundle a linear model was developed by using the dependent variables, such as the fast neutron fluences and the bundle coolant temperatures. The training data set was selected using the subtractive clustering method. The data of 39 channels that consist of 80 percent of a total of 49 measured channels from Units 2, 3, and 4 of the Wolsung nuclear plant in Korea were used to develop the BPLM. The data from the remaining 10 channels were used to test the developed BPLM. The BPLM was optimized by the maximum likelihood estimation method. The developed BPLM to predict PT diametral creep was verified using the operating data gathered from Units 2, 3, and 4. Two error components for the BPLM, which are the epistemic error and the aleatory error, were generated. The diametral creep prediction and two error components will be used for the generation of the regional overpower trip setpoint at the corresponding effective full power days. The root mean square (RMS) errors were also generated and compared to those from the current prediction method. The RMS errors were found to be less than the previous errors.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.537-538
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• 2004

The performance of the ECWS was predicted for the anticipated operational occurrences. The inadvertent close of loop isolation valve is the most severe case for the five anticipated operational occurrences considered in this design and meets the design criteria of the ECWS. The correlation of critical heat flux for the geometry of three pins sub-channel analysis will be studied in the feature.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 2003년도 춘계 학술발표회 논문집
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• pp.415-423
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• 2003

Korea Electric Power Research Institute(KEPRI) had worked a comprehensive Plant Lifetime Management (PLiM) project for a CANDU plant in corporation with Korea Hydro and Nuclear Power(KHNP). The project had been performed to understand the aging status of major components screened from the plant and to address provisions for the continued operation over its design life. A feasibility of the continued operation was reviewed in the aspects of technology, economics, and regulatory environments. This paper introduces general approach of aging assessment, screening of critical components and an experience of aging assessment for an example of fuel channel that is the most critical component in CANDU plant.

• Nuclear Engineering and Technology
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• 제36권2호
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• pp.165-174
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• 2004

The distributed resistance model has been recently implemented into the MATRA-LMR code in order to improve its prediction capability over the wire-wrap model for a flow blockage analysis in the LMR. The code capability has been investigated using experimental data observed in the FFM (Fuel Failure Mock-up)-2A and 5B for two typical flow conditions in a blocked channel. The predicted results by the MATRA-LMR with a distributed resistance model agreed well with the experimental data for wire-wrapped subchannels. However, it is suggested that the parameter n in the distributed resistance model needs to be calibrated accurately for a reasonable prediction of the temperature field under a low flow condition. Finally, the analyses of a blockage for the assembly of the KALIMER design are performed. Satisfactory results by the MATRA-LMR code were obtained through and rerified a comparison with results of the SABRE code.

• Nuclear Engineering and Technology
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• 제52권1호
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• pp.51-58
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• 2020

Despite the recent growth of interest in molten salt reactor technology and the crucial role which heat transfer plays in the design of power reactors, specific studies on the design of heat exchangers for the Molten Salt Fast Reactor have not yet been performed. In this work we deliver a preliminary but quantitative analysis of the intermediate heat exchangers, based on reference design data from the SAMOFAR H2020-Euratom project. Two different promising reference technologies are selected for study thanks to their compactness features, the Printed Circuit and the Helical Coil heat exchangers. We present preliminary design results for each technology, based on simplified design tools. Results highlight the limiting effects of the compactness constraints imposed on the fuel salt inventory and the allowed size. Large pressure drops on both flow sides are to be expected, with negative consequences on pumping power and natural circulation capabilities. The small size required for the flow channels also represents possible fabrication issues and safety concerns regarding channel blockage.

• Nuclear Engineering and Technology
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• 제26권4호
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• pp.570-577
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• 1994

The loss of water in the carbon steel balls and water region of the end shield for CANDU 6 reactor could lead to significant temperature gradient through the end shield structure which amy result in the excessive deformation. With an assumed end shield drained scenario, the heat deposition rates were calculated through the end shield associated with the central fuel channel during full power operation as an initial step to thermal stress analysis. The drained case was compared with that of water present normal case in therms of heat deposition rater and the total heating throughout the end shield regions. The compared results show that the heat deposition and the total heating remain almost the same between the two cases. It was found that the change of volume integrated flux in the end shield regions due to the loss of water contribute a negligible effect on the heat deposition in this region.

• Nuclear Engineering and Technology
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• 제52권1호
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• pp.19-26
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• 2020

Small nuclear reactor features higher power capacity, longer operation life than conventional power sources. It could be an ideal alternative of existing power source applied for special equipment for terrestrial or underwater missions. In this paper, a 25kWe heat pipe cooled reactor power source applied for multiple use is preliminary designed. Based on the design, a thermal-hydraulic analysis code for heat pipe cooled reactor is developed to analyze steady and transient performance of the designed nuclear reactor. For reactor design, UN fuel with 65% enrichment and potassium heat pipes are adopted in the reactor core. Tungsten and LiH are adopted as radiation shield on both sides of the reactor core. The reactor is controlled by 6 control drums with B₄C neutron absorbers. Thermoelectric generator (TEG) converts fission heat into electricity. Cooling water removes waste heat out of the reactor. The thermal-hydraulic characteristics of heat pipes are simulated using thermal resistance network method. Thermal parameters of steady and transient conditions, such as the temperature distribution of every key components are obtained. Then the postulated reactor accidents for heat pipe cooled reactor, including power variation, single heat pipe failure and cooling channel blockage, are analyzed and evaluated. Results show that all the designed parameters satisfy the safety requirements. This work could provide reference to the design and application of the heat pipe cooled nuclear power source.

• 한국마린엔지니어링학회:학술대회논문집
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• 한국마린엔지니어링학회 2005년도 전기학술대회논문집
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• pp.1083-1090
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• 2005

Repeated ribs are used on heat exchange surfaces to promote turbulence and enhance convective heat transfer. Applications include fuel rods of gas-cooled nuclear reactors, inside cavities of turbine blades, and internal surfaces pipes used in heat exchangers. Despite the great number of literature papers, only few experimental data concern detailed distributions of friction factors and heat transfer coefficients in square channels varying the number of rough walls. This issue is tackled by investigating effects of different number of ribbed walls on heat transfer and friction characteristics in square channel. The rough wall have a 45$^{\circ}$ inclined square rib. Uniform heat flux is maintained on whole inner heat transfer channel area. The heat transfer coefficient and friction factor values increase with increasing the number of rough walls.

• 동력기계공학회지
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• 제9권1호
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• pp.30-35
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• 2005

Repeated ribs are used on heat exchanger surfaces to promote turbulence and to enhance convective heat transfer. Applications include fuel rods of gas-cooled nuclear reactors, inside cavities of turbine blades, and internal surfaces pipes used in heat exchangers. Despite the great number of literature papers, only few experimental data concerns detailed distributions of friction factors and heat transfer coefficients in square channels varying the number of rough walls. This issue was tackled by investigating effects of different number of ribbed walls on heat transfer and friction characteristics in square channel. The rough wall had a $45^{\circ}$ inclined square rib. Uniform heat flux was maintained on the whole inner heat transfer channel area. The heat transfer coefficient and friction factor values increased with increasing the number of rough walls.

• 대한기계학회논문집B
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• 제29권3호
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• pp.340-348
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• 2005

Repeated ribs are used on heat exchange surfaces to promote turbulence and enhance convective heat transfer. Applications include fuel rods of gas-cooled nuclear reactors, inside cavities of turbine blades, and internal surfaces pipes used in heat exchangers. Despite the great number of literature papers, only few experimental data concern detailed distributions of friction factors and heat transfer coefficients in square channels varying the number of rough walls. This issue is tackled by investigating effects of different number of ribbed walls on heat transfer and friction characteristics in square channel. The rough wall have a $45{\circ}C$ inclined square rib. Uniform heat flux is maintained on whole inner heat transfer channel area. The heat transfer coefficient and friction factor values increase with increasing the number of rough walls.