• The Transactions of The Korean Institute of Electrical Engineers
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• v.59 no.8
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• pp.1456-1463
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• 2010

The safety critical systems in nuclear power plants should be designed to have a high level of fault tolerant capability because those systems are used for protection or mitigation of the postulated accidents of nuclear reactor. Due to increasing of the system complexity of the digital based system in nuclear fields, the reliability of the digital based systems without an auto-test or a self-diagnostic feature is generally lower than those of analog system. To overcome this problem, additional redundant architectures in each redundant channel and self-diagnostic features are commonly integrated into the digital safety systems. The self diagnostic function is a key factor for increasing fault tolerant capabilities in the digital based safety system. This paper presents an availability and safety evaluation model to analyze the effect to the system's fault tolerant capabilities depending on self-diagnostic features when the loss or erroneous behaviors of self-diagnostic function are expected to occur. The analysis result of the proposed model on the several modules of a safety platform shows that the improvement effect on unavailability of each module has generally become smaller than the result of usage of conventional models and the unavailability itself has changed significantly depending on the characteristics of failures or errors of self-diagnostic function.

• Journal of Energy Engineering
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• v.23 no.3
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• pp.96-101
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• 2014

The Instrument Channel setpoints of the Reactor Protection System(RPS) and the Engineered Safety Feature Actuation System(ESFAS) ensures the safety of Nuclear Power Plants (NPPs), and the actuation of the protection system should be guaranteed on power change condition. The goal of this study is to verify the appropriateness of the sensor drift and rack drift which are important factors for setpoints evaluation and to improve the setpoints margin using the operation data, design specifications and operation manuals of the NPPS.

• Journal of the Korean Institute of Illuminating and Electrical Installation Engineers
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• v.25 no.4
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• pp.71-78
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• 2011

This paper contains the selection of operation parameters for protection and control of steam turbine in a Korea Standard Type Nuclear Power Plant. The safety of nuclear reactor must be ensured which generates nuclear energy and produces steam. Also, the safety of turbine, which consume the nuclear energy as a core machine, must be ensured. For the purpose of this, we describe how the operating parameters were selected, reviewed, implemented into the operator console and finally put into actual operation of the system.

• Proceedings of the Korean Institute of Industrial Safety Conference
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• 2003.05a
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• pp.102-107
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• 2003

원자력발전소의 산업계와 규제기관에서는 원자력발전소의 성능(performance)수준 향상과 대 국민 알권리 충족, 규제업무 효율화를 위하여 정량적인 성능지표를 개발하여 사용해왔다. 국내의 원자력 규제 기관인 한국원자력안전기술원은 WANO (world association of nuclear operators)와 미국 핵규제위원회(NRC)의 발전소 종합성능평가 (reactor oversight process)에서 사용하는 성능지표를 토대로 국내 원자력발전소 성능지표를 개발하여 사용 중이다 ( 1 ). 개발된 국내 원자력발전소의 성능지표는 초기사건, 안전계통 신뢰도, 방사선 안전 등 11개 지표로 이루어졌다. 각 지표별 등급을 4개(녹색, 하늘색, 노랑색, 주황색)로 나누어 알기 쉽게 표시하였다.(중략)

• Journal of the Korean Society of Safety
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• v.21 no.4 s.76
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• pp.49-54
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• 2006

The evaluation of elastic-plastic fracture characteristic was investigated in ferrite steel SA 516- Gr70 used for reactor coolant piping elbow and support skirt of pressure vessels. This paper describes the effect of temperature on J-R curve characteristic of this material. The elastic-plastic fracture mechanics parameter J is obtained with unloading compliance method. The test method were analyzed according to ASTM E 813-89 and E 1152-89. Unloading compliance $J_{IC}$ tests were performed on 1 CT specimens at varied temperatures from $25^{\circ}C$ to about $400^{\circ}C$ using a high temperature extensometer. At all temperature, valid $J_{IC}$ measurements could be made and $J_{IC}$ decreased with increasing temperature. SEM fractography schematically illustrates microvoid initiation, growth and coalescence at the tip of a preexisting crack.

• Journal of the Korean Professional Engineers Association
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• v.43 no.2
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• pp.45-52
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• 2010

At the end of last year, Korean nuclear power plants, APR-1400 and a research reactor have been contracted to build plants at United Arab Emirates and Jordan. Since 1959, a historical background of nuclear technologies in Korea is summarized. The safety requirements for Instrumentation and Control (I&C) systems in Nuclear Power Plants (NPP) are discussed. Specific descriptions on the typical safety classification of I&C systems, the definitions of the electrical class 1E and the countermeasures against common caused failures are provided. And summaries of typical I&C systems such as the protection systems, the control systems, the instrumentations, the monitoring systems and a control room in NPP are introduced. Strict requirements on the development of the digital computer systems in nuclear applications are described.

• Proceedings of the KSME Conference
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• 2002.08a
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• pp.51-54
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• 2002

When a cold HPSI (High Pressure Safety Injection) fluid associated with an design basis accident, such as LOCA (Loss of Coolant Accident), enters the cold legs of a stagnated primary coolant loop, thermal stratification phenomena will arise due to incomplete mixing. If the stratified flow enters a reactor pressure vessel downcomer, severe thermal stresses are created in a radiation embrittled vessel wall by local overcooling. Previous thermal-mixing analyses have assumed that the thermal stratification phenomena generated in stagnated loop of a partially stagnated coolant loop are neutralized in the vessel downcomer by strong flow from unstagnated loop. On the basis of these reasons, this paper presents the thermal-mixing analysis results in order to identify the fact that the cold plume generated in the vessel downcomer due to the thermal stratification phenomena of the stagnated loop is affected by the strong flow of the unstagnated loop.

• Journal of Power System Engineering
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• v.11 no.3
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• pp.3-8
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• 2007

The emergency AC power supply system of the nuclear power plant is designed to supply the power to the nuclear reactor at the emergency operating condition. The safety function of the diesel generator at the nuclear power plant is to supply AC electric power to the plant safety system whenever the preferred AC power supply is unavailable. The reliable operation of onsite emergency diesel generator should be ensured by a conditioning monitoring system designed to maintain and monitor and forecast the reliability level of diesel generator. To do this kind of diesel generator condition monitoring we reviewed several operating factors and history of the wolsong unit 3 diesel generator and selected the proper conditioning monitoring operating factors.

• Journal of IKEEE
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• v.18 no.4
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• pp.566-571
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• 2014

A safety class field programmable gate array based system in nuclear power plant has been developed to improve the diversity. Testbench is necessary to satisfy the technical reference, IEC-62566, for verification and validation of register transfer level code. We use the open verification methodology(OVM) developed by standard body. We show that our testbench can use random input for test. And also we show that reusability of block level testbench for the integration level testbench, which is very efficient for large scale system like nuclear reactor protection system.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.61 no.3
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• pp.367-372
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• 2012

In nuclear plants, UVR (under voltage relay, 27r) of 1E bus, which protected and supplied power to essential loads, to safety trip of reactor and supplied to starting signal of EDG (emergency diesel generators) automatically. therefore UVR tolerances setting and calculation method has been important to nuclear facility. If calculation and tolerances values differ or ignore, may induced power loss and economical loss by protective failure. This paper show results for calculation methods, and whether dependant or independent methods for factors. included whether PT (potential transformer/voltage transformer) tolerance or not adapted, and based on UVR setting method within a difference minimum and maximum of rated voltage to safety operation in nuclear plants.