• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.1291-1292
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.419-422
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• 1995

The Atmospheric Dump Valve (ADV) and Turbine Bypass Valve (TBV) capacity test was performed at 30% power level during the YGN 3 Power Ascension Test period. In this test, several plant data were measured to calculate the ADV and TBV capacity considering that critical condition was developed through the ADV and TBV. The test results show that the test acceptance criteria are met.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.403-408
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• 1996

The subchannel analysis was performed to verify the behavior of hot channel characteristics and obtain the information to support the core thermal-hydraulic behavior at post-trip steam line break with low flow condition. During this postulated accident, buoyancy-induced cross flow occurs, and the coupled nuclear and thermal-hydraulic interactions become important. The code predictions with TORC are in good agreement with the test data. Under such conditions, the mass flow increase in the hot channel by buoyancy-induced cross flow depends on the parameter $GR^{*}\;/\;Re^2$, and buoyancy effect becomes more noticeable as $GR^{*}\;/\;Re^2$ increases.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.155.2-155
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• 2002

A visual environment for system analysis codes (hereinafter called "ViSA") has been developed to support code users in their input preparations, code executions, and output interpretations. ViSA provides a more convenient way for base input data generation and modification on a user-friendly basis. It also provides on-line graphical displays to give an in-depth understanding of transient thermal-hydraulic behaviors in nuclear power plants. This paper presents the main features of ViSA.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.61-66
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• 1996

High pressure. high temperature steam/water blowdown test loop has been constructed. The loop simulates a pressurizer. depressurizalion system and In-Containment Refueling Water Storage Tank (IRWST) with full pressure and temperature conditions. and will be used to generate data for development of an optimal sparser as well as for design of safety/automatic depressurization system. In addition. experiments for reactor safety and pressurizer thermal hydraulics are scheduled. In this paper. general description of the Blowdown and Condensation (B&C) Loop will be given together with the test program.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.588-595
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• 1995

The Load Rejection to House Load test at 100% power was successfully performed during the YGN 4 PAT period. In this test, all plant control systems automatically controlled the plant from 100% power to house load operation mode. The LTC code, which was used in the performance analysis during the design process of YGN 3&4, predictions of the test agreed with the measured data demonstrating the validity of the code as well as the completeness of the plant design.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.13-14
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• 2005

KAERI is performing a project on out-pile test technology development for a full scale PWR fuel assembly. We have introduced the hydraulic test facility, a test assembly, test parameters, test methods, and a data acquisition system. The start up test will be in the middle of March 2005 and the main test will be accomplished by the end of 2006. The established test facility and measuring technique will contribute to the satisfaction of domestic needs for the design verification to improve the reliability of a PWR plant operation.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.483-488
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• 1998

A pressure drop model for the PWR grids with and without mixing device is proposed at single phase based on the fluid mechanistic approach. Total pressure loss is expressed in additive way for form and frictional losses. The general friction factor correlations and form drag coefficients available in the open literatures are used to the model. As the results, the model shows better predictions than the existing ones for the non-mixing grids, and reasonable agreements with the available experimental data for mixing grids. Therefore it is concluded that the proposed model for pressure drop can provide sufficiently good approximation for grid optimization and design calculation in advanced grid development.

• Nuclear Engineering and Technology
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• v.30 no.4
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• pp.298-307
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• 1998

In order to design mutual communication between a distributed control system and a function test facility, we used the Inter-Process Communication(IPC) in two systems and Transmission Control Protocol/Internet Protocol(TCP/IP) protocol. The data from the function test facility are put in the shared memory using an IPC, which is then accessed by the distributed control system through an Application Program Interface(API). The server in the function test facility includes two processes(one for sending and one for receiving), which are generated by the fork function from the client signal. The client in the distributed control system includes two separate programs(one for receiving and one for sending).

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.296-302
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• 1996

The Loss of a Main Feedwater Pump test at 100% Power for YGN 4 was simulated in order to verify and validate the KSNPA. The comparison of the test data with the KSNPA prediction results showed reasonable agreement in the trends of the major plant parameters. All plant control systems including NSSS and T/G control systems are properly actuated and stabilized the plant conditions to a new steady state conditions in the KSNPA. From the comparison results, the KSNPA showed its capability to simulate the LOMFP event for the Korean Standard Nuclear Power Plant.