• Journal of the Korean Society for Aeronautical & Space Sciences
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• v.32 no.4
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• pp.67-72
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• 2004

A regenerative cooling system has been designed through empirical 1-D analysis for a liquid rocket engine of 30-ton-level thrust. The hot-gas-side wall temperature from 1-D analysis shows 100K difference compared to 3D CFD analysis. Two variations of design with same cooling performance are suggested for different maximum channel widths i.e., 4mm and 2mm. The coolant pressure drop of the latter design is higher by 20%. The maximum liner temperature is about 700K when TBC and the thermal resistance of carbon deposit are considered. So film cooling is recommended to increase the cooling capacity as the present cooling capacity is insufficient

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.1045-1052
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• 2009

A calculation model is developed to predict the transient free surface flow on the containment floor following a loss-of-coolant accident (LOCA) of pressurized water reactors (PWR) for the use of debris transport evaluation. The model solves the two-dimensional Shallow Water Equation (SWE) using a finite volume method (FVM) with unstructured triangular meshes. The numerical scheme is based on a fully explicit predictor-corrector method to achieve a fast-running capability and numerical accuracy. The Harten-Lax-van Leer (HLL) scheme is used to reserve a shock-capturing capability in determining the convective flux term at the cell interface where the dry-to-wet changing proceeds. An experiment simulating a sudden break of a water reservoir with L-shape open channel is calculated for validation of the present model. It is shown that the present model agrees well with the experiment data, thus it can be justified for the free surface flow with accuracy. From the calculation of flow field over the simplified containment floor of APR1400, the important phenomena of free surface flow including propagations and interactions of waves generated by local water level distribution and reflection with a solid wall are found and the transient flow rates entering the Holdup Volume Tank (HVT) are obtained within a practical computational resource.

• Progress in Superconductivity
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• v.8 no.2
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• pp.158-163
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• 2007

YBCO do superconducting quantum interference device (SQUID) magnetometers based on bicrystal junctions have been fabricated for magnetocardiograph (MCG) measurements. We could fabricate YBCO SQUID magnetometers having magnetic field noise of about $20fT/Hz^{1/2}$ at white noise region. We have developed an MCG system employing the high performance SQUID magnetometers. The lightweight MCG system, requiring liquid nitrogen as a coolant, consists of 6-channel SQUID sensors, an adjustable patient bed with sliding motion, and data analyses software. The MCG system could record quite clear MCG signals in a room with moderate magnetic shielding. In normal operation with multi-position MCG measurements, we could obtain clear 48-point mappings of magnetic field map and current source map with high enough signal qualities far clinical trials.

• Journal of the Korean Society for Aeronautical & Space Sciences
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• v.39 no.6
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• pp.559-566
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• 2011

This paper presents the flow passage design for a window cooling device, which have a conical poppet valve and an emissive orifice. Computational flow analysis and experiment are conducted according to the poppet strokes. The results show satisfactory flow characteristics that pressure is reduced enough to endure material strength and the flow does not choked inside window. The correction factor of discharge coefficients is found between 2-dimensional analysis and experiments, which is applied to control coolant flow rates of the window cooling device.

• Nuclear Engineering and Technology
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• v.33 no.2
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• pp.209-224
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• 2001

The SSC-K system analysis code is under development at the Korea Atomic Energy Research Institute (KAERI) as a part of the KALIMER project. The SSC-K code is being used as the principal tool for analyzing a variety of off-normal conditions or accidents of the preliminary KALIMER design. The SSC-K code features a multiple-channel core representation coupled with a point kinetics model with reactivity feedback. It provides a detailed, one-dimensional thermal-hydraulic simulation of the primary and secondary sodium coolant circuits, as well as the balance-of-plant steam/water circuit. Recently a two-dimensional hot pool model was incorporated into SSC-K for analysis of thermal stratification phenomena in the hot pool. In addition, SSC-K contains detailed models for the passive decay heat removal system and a generalized plant control system. The SSC-K code has also been applied to the computational engine for an interactive simulation of the KALIMER plant. This paper presents an overview of the recent activities concerned with SSC-K code model development This paper focuses on both descriptions of the newly adopted thermal hydraulic and neutronic models, and applications to KALIMER analyses for typical anticipated transients without scram.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.581-588
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• 2013

This paper deals with the Safety Analysis for $CANDU^{(R)}$ 6 nuclear reactors as affected by main Heat Transport System (HTS) aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermal-hydraulic analytic models. Safety analyses ranging from the assessment of safety limits associated with the prevention of intermittent fuel sheath dryout for a slow Loss of Regulation (LOR) analysis and fission gas release after a fuel failure are summarized. Specifically for fission gas release, the thermal-hydraulic analysis for a fresh core and an 11 Effective Full Power Years (EFPY) aged core was summarized, leading to the most severe stagnation break sizes for the inlet feeder break and the channel failure time. Associated coolant conditions provide the input data for fuel analyses. Based on the thermal-hydraulic data, the fission product inventory under normal operating conditions may be calculated for both fresh and aged cores, and the fission gas release may be evaluated during the transient. This analysis plays a major role in determining possible radiation doses to the public after postulated accidents have occurred.

• Nuclear Engineering and Technology
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• v.44 no.4
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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.

• Corrosion Science and Technology
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• v.15 no.5
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• pp.245-252
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• 2016

The pipelines and equipments are degraded by flow-accelerated corrosion (FAC), and a large-scale test facility was constructed for simulate the FAC phenomena in secondary coolant environment of PWR type nuclear power plants. Using this facility, FAC test was performed on weld pipe (carbon steel and low alloy steel) at the conditions of high velocity flow (> 10 m/s). Wall thickness was measured by high temperature ultrasonic monitoring systems (four-channel buffer rod type and waveguide type) during test period and room temperature manual ultrasonic method before and after test period. This work deals with the complex effects of flow velocity on the wall thinning in weld pipe and the test results showed that the higher flow velocity induced different increasement of wall thinning rate for the carbon steel and low alloy steel pipe.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.124-129
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• 1998

A sensitivity study has been done to determine the composition of DUPIC fuel from the viewpoint of neutronics fuel design. The spent PWR fuel compositions were generated and fissile contents adjusted by blending fresh uranium after mixing two spent PWR fuel assemblies. The $^{239}$ Pu and $^{235}$ U enrichments of DUPIC fuel were adjusted by controlling the amount of fresh uranium feed and the ratio of slightly enriched and depleted uranium in the fled uranium. Based on the material balance calculation, it is recommended that DUPIC fuel composition be such that spent PWR fuel utilization is more than 90%.. A sensitivity study on the temperature reactivity coefficient of DUPIC fuel has shown that it is desirable to increase the $^{239}$ Pu and $^{235}$ U contents to reduce both the fuel and coolant temperature coefficients. On the other hand, refueling simulations of the DUPIC core have shown that the channel power peaking factor, which is a measure of the reactor trip margin, increases with the total fissile content. Considering these neutronic characteristics of the DUPIC fuel, il is recommended to have enrichments of 0.45 and 1.00 wt% for $^{239}$ Pu and $^{235}$ U, respectively.

• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
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• 1999.05a
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• pp.434-437
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• 1999

An EM pump is used for the purpose of transporting the electrically conducting liquid sodium of the high temperature that is used as a coolant in the liquid metal reactor. In the present study, the pilot pump has been designed and manufactured for the high temperature of $600^{\circ}C$ by the equivalent circuit materials and the consideration of the materials and functions. The length and diameter of the pump are given as 84 cm and 10 cm each due to the fixed geometry of the circulation system to be installed. The characteristic of the developing pressure and efficiency is found out by using Laithewaite\`s standard design formula. It is shown that the developing pressure and efficiency are maximized at the frequency of 15 Hz from the curve. The annular channel gap of 3.95 mm is selected in the range of the reasonable hydraulic frictional loss. The components of the pump consist of the material for the high temperature. And then, the pump is manufactured to have the nominal flowrate of 40 1/min and developing Pressure of 1.3 bar.