• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1994.11a
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• pp.135-140
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• 1994

An improved steam generator model has been developed for the DSNP simulation of normal operational transient behavior of CANDU nuclear power plant. For more realistic prediction of steam generator behavior during transient, tube bundle region is divided into two separate control volumes, subcooled region and saturated region, and the variation of thermal hydraulic properties in the control volume is accounted for more realistic estimates of outlet enthalpy of each control volume. Test results for typical CANDU operational transient case show reasonable transient behavior of steam generator with overall CANDU operation and improved operational characteristics of steam generator with power variation.

• Proceedings of the KIEE Conference
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• 1998.07g
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• pp.2284-2289
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• 1998

The Rotor Stress Indicator is an integrated system of hardware and program components which has been designed to read an assortment of turbine temperature and speed input devices, perform an analysis of the temperature induced stresses and output pertinent temperature and stress information to guide the turbine operator during turbine prewarming, start-ups, load changes, and shut-downs. The purpose of the RSI is to provide guidance to the plant operator during startup, shutdown, loading, and unloading of the turbine. Since the stresses are a function of the temperature changes to which the turbine is exposed, the RSI also provides guidance for operation of the boiler main steam and reheat steam temperatures as they affect the rotor stresses. This may permit more efficient overall boiler turbine start-ups. In this paper, new rotor stress analysis algorithm for RSI is introduced and compared with present system which has been used in thermal power plant.

• Journal of Power System Engineering
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• v.8 no.3
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• pp.18-22
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• 2004

Boiler is one of the most important utilities providing steam to turbine in order to supply mechanical energy in thermal power plant. It is composed of thousands of tubes for high efficient heat transfer. The material for boiler tubes is used in such high temperature and pressure condition as $540^{\circ}C$, 22MPa. The boiler tube material is required to resist creep damage, fatigue cracking, and corrosion damages. 2.25%Cr-1Mo steel is used for conventional boiler tubes, and austenitenite stainless steel is used for higher temperature boiler tubes. But the temperature and pressure of steam in power plant became higher for high plant efficiency. So, the property of boiler tube material must be upgaded to fit the plant property. Several boiler tube material was developed to fit such conditions. X20CrMoV12.1 steel is also developed in 1980's and used for superheater and reheater tubes in supercritical boilers. The material has martensite microstructures which is difficult to evaluate the degradation. In this thesis, degrade the X20CrMoV12.1 steel at high temperatures in electric furnace, and evaluate hardness with Vickers hardness tester and strengths with Indentation tester.

• Journal of Ocean Engineering and Technology
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• v.13 no.2 s.32
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• pp.90-98
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• 1999

Boiler high-temperature pipelines such as main steam pipe, header and steam drum in fossil power plants are degraded by creep damage due to severe operationg conditions like high temperature and high pressure for an extended period time. Such material degradation lead to various component faliures causing serious accidents at the plant. Conventional measurement techniques such as replica method, electric resistance method, and hardness test method have such disadvantages as complex preparation and measurement procedures, too many control parameters, and therefore, low practicality and they were applied only to component surfaces with good accessibility. In this study, both artificial creep degradation test using life prediction formula and frequency analysis by ultrasonic tests for their preparing creep degraded specimens have been carried out for the purpose of nondestructive evaluation for creep damage which can occur in high-temperature pipelline of fossil power plant. As a result of ultrasonic tests for crept specimens, we confirmed that the high frequency side spectra decrease and central frequency components shift to low frequency bans, and bandwiths decrease as increasing creep damage in backwall echoes.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.3
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• pp.7-12
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• 2012

Steam generator tubes are critical boundary of the primary and secondary side in nuclear power plants. Eddy current testing is commonly used as the method of non-destructive testing for the safety and integrity of steam generator tubes in the nuclear power plants. Changes in the geometric shape act as a stress concentration factor likely to cause a defect during the steam generator operation. The mixed-signals with the geometric shape are distorted and attributes that are difficult to detect signals. An example is bending stress due to compression process at a U-bend occurring in the intrados region which has a small radius of curvature. The resulting change in the geometric shape may lead to a dent like occurrences. The dent can cause stress concentration and generates stress corrosion cracks. In this study, the steam generator tubes of nuclear power plant were selected to study for analysis of mixed-signal containing dent and stress corrosion cracks.

• Proceedings of the KIEE Conference
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• 2007.07a
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• pp.1624-1625
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• 2007

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.2
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• pp.77-87
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• 2018

In the case of high-energy line breaks in nuclear power plants, supersonic steam jet is formed due to the rapid depressurization. The steam jet can cause impingement load on the adjacent structures, piping systems and components. In order to secure the design integrity of the nuclear power plant, it is necessary to evaluate the load characteristics of the steam jet generated by high-energy pipe rupture. In the design process of nuclear power plant, jet impingement load evaluation was usually performed based on ANSI/ANS 58.2. However, U.S. NRC recently pointed out that ANSI/ANS 58.2 oversimplifies the jet behavior and that some assumptions are non-conservative. In addition, it is recommended that dynamic analysis techniques should be applied to consider transient load characteristics. Therefore, it is necessary to establish an evaluation methodology that can analyze the dynamic load characteristics of steam jet ejected when high energy pipe breaks. This research group has developed and validated the CFD analysis methodology to evaluate the transient behavior of supersonic impinging jet in the previous study. In this study, numerical study on the transient load characteristics of supersonic steam jet impingement was carried out and amplitude and frequency analysis of transient jet load was performed.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.715-720
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• 2001

The main reason for applying positive pressure variable clearance packing in fossil power plant is high efficiency and energy saving movement in the government. This study intends to analyze the turbine efficiency through the shaft packing improvement in thermal power plant and makes its comparison to that of the used packing.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.31 no.2 s.257
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• pp.277-284
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• 2007

The steam generator in a nuclear power plant is a large heat exchanger that uses heat from a reactor to generate steam to drive the turbine generator. Rupture of a steam generator tube can result in release of fission products to environment outside. Therefore, an accurate integrity assessment of the steam generator tubes with cracks is of great importance for maintaining the safety of a nuclear power plant. The steam generator tubes are supported at regular intervals by support plates and rotations of the tubes are restrained. Although it has been reported that the limit load for a circumferential crack is significantly affected by boundary condition of the tube, existing limit load solutions do not consider the restraining effect of support plate correctly. In addition, there are no limit load solutions for circumferential cracks in U-bend region with the effect of the support plate. This paper provides detailed limit load solutions for circumferential cracks in top of tube sheet and the U-bend regions of the steam generator tube with the actual boundary conditions to simulate the restraining effect of the support plate. Such solutions are developed based on three dimensional finite element analyses. The resulting limit load solutions are given in a polynomial form, and thus can be simply used in practical integrity assessment of the steam generator tubes.

• Journal of the Korean Society for Nondestructive Testing
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• v.32 no.5
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• pp.545-550
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• 2012

Steam generator is one of the major components of nuclear power plant and steam generator tubes are the pressure boundary between primary and secondary side, which makes them critical for nuclear safety. As the operating time of nuclear power plant increases, not only damage mechanisms but also scaled deposits on steam generator tubes are known to be problematic causing tube support flow hole blockage and heat fouling. The ability to assess the extent and location of scaled deposits on tubes became essential for management and maintenance of steam generator and eddy current bobbin data can be utilized to measure thickness of scale on tubes. In this paper, tube reference standards with various thickness of scaled deposit has been set up to provide information about the overall deposit condition of steam generator tubes, providing essential tool for steam generator management and maintenance to predict and prevent future damages. Also, methodology to automatically measure scale thickness on tubes has been developed and applied to field data to estimate overall scale amount.