• 대한전기학회:학술대회논문집
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• 대한전기학회 1995년도 추계학술대회 논문집 학회본부
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• pp.175-177
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• 1995

The sensor signal validation and failure detection system must be able to detect, isolate, and identify sensor degradation as well as provide a reconstruction of the measurements. In this study, this is accomplished by combining the neural network, the Generalized Consistency Checking(GCC), and the Sequential Probability Ratio Test(SPRT) method in a decision estimator module. The GCC method is a computationally efficient system for redundant sensors, while the SPRT provides the ability to make decisions based on the degradation history of a sensor. The methodology is also extended to the detection of noise degradation. The acceptability of the proposed method is demonstration by using the simulation data in safety injection accident of nuclear power plants. The results show that the signal validation and sensor failure detection system is able to detect and isolate a bias failure and noise type failures under transient conditions. And also, the system is able to provide the validated signal by reconstructing the measurement signals in the failure conditions considered.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회A
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• pp.1917-1921
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• 2008

Ultraviolet-nanoimprint lithography (UV-NIL) is promising technology for cost effectively defining micro/nano scale structure at room temperature and low pressure. In addition, this technology is fascinating because of it's possibility for high-throughput patterning without complex processes. However, to acquire good micro/nano patterns using this technology, there are some challenges such as uniformity and fidelity of patterns, etc. In this paper, we have focused on uniform contact mechanism and performed contact mechanics analysis. The dimension of the flexible sheet to get adequate uniform contact area has been obtained from contact mechanics simulation. Based on this analysis, we have made a uniform pressurizing device and confirmed its uniform pressurized zone using a pressure sensing paper.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회A
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• pp.85-87
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• 2008

DMWs are common feature of the PWR in the welded connections between carbon steel and stainless steel piping. The nickel-based weld metal, Alloy 82/182, is used for welding the dissimilar metals and is known to be susceptible to PWSCC. A round-robin program has been implemented to benchmark the numerical simulation of the transient temperature and weld residual stresses in the DMWs. To solve the round-robin problem related to Pressurizer Safety & Relief nozzle, the thermal elasto-plastic analysis is performed in the DMW by using the FEM. The welding includes both the DMW of the nozzle to safe-end and the SMW of the safe-end and piping. Major results of the analyses are discussed: The axial and circumferential residual stresses are found to be -88MPa(225MPa) and -38MPa(293MPa) on the inner surface of the DMW; where the values in parenthesis are the residual stresses after the DMW. Thermo-mechanical interaction by the SMW has a significant effect on the residual stress fields in the DMW.

• 한국전산유체공학회지
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• 제14권1호
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• pp.78-85
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• 2009

A combined theoretical and numerical study is conducted to design a slit nozzle for large-area liquid coating. The objectives are to guarantee the uniformity in the injected flow and to provide the capability of explicit control of flow rate. The woking fluid is a dilute aqueous solution containing single-walled carbon nanotubes and its low viscosity and the presence of dispersed materials pose technical hurdles. A theoretical analysis leads to a guideline for the geometric design of a slit nozzle. The CFD-based numerical experiment is employed as a verification tool. A new flow passage unit, connected to the nozzle chamber, is proposed to permit the control of flow rate by using the commodity pressurizer. The numerical results confirm the feasibility of this idea. The optimal geometry of internal structure of the nozzle has been searched for numerically and the related issues are discussed.

• 대한기계학회논문집A
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• 제32권10호
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• pp.873-881
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• 2008

Weld overlay is one of the residual stress mitigation methods which arrest crack initiation and crack growth. Therefore weld overlay can be applied to the region where cracking is likely to be. An overlay weld used in this manner is termed a preemptive weld overlay(PWOL). In pressurized water reactor(PWR) dissimilar metal weld is susceptible region for primary water stress corrosion cracking(PWSCC). In order to examine the effect of PWOL on residual stress mitigation, PWOL was applied to a specific dissimilar metal weld of Kori nuclear power plant by finite element analysis method. As a result, strong compressive residual stress was made in PWSCC susceptible region and PWOL was proved effective preemptive repair method for weldment.

• Journal of Mechanical Science and Technology
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• 제19권5호
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• pp.1206-1215
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• 2005

The thermal stratification phenomena, frequently occurring in the component of nuclear power plant system such as pressurizer surge line, steam generator inlet nozzle, safety injection system (SIS), and chemical and volume control system (CVCS), can cause through-wall cracks, thermal fatigue, unexpected piping displacement and dislocation, and pipe support damage. The phenomenon is one of the unaccounted load in the design stage. However, the load have been found to be serious as nuclear power plant operation experience accumulates. In particular, the thermal stratification by the turbulent penetration or valve leak in the SIS and SCS pipe line can lead these safety systems to failure by the thermal fatigue. Therefore in this study an 1/10 scaledowned experimental rig had been designed and installed. And a series of experimental works had been executed to measure the temperature distribution (thermal stratification) in these systems by the turbulent penetration, valve leak, and heat transfer through valve. The results provide very valuable informations such as turbulent penetration depth, the possibility of thermal stratification by the heat transfer through valve, etc. Also the results are expected to be useful to understand the thermal stratification in these systems, establish the thermal strati­fication criteria and validate the calculation results by CFD Codes such as Fluent, Phenix, CFX.

• Nuclear Engineering and Technology
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• 제35권3호
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• pp.179-190
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• 2003

The multiple steam generator tube rupture (SGTR) event scenario with available safety systems was experimentally and analytically evaluated. The experiment was conducted on the large scaled test facility to simulate the multiple SGTR event and investigate the effectiveness of operator actions. As a result, it indicated that the opening of pressurizer power operated relief valve was significantly effective in quickly terminating the primary-to-secondary break flow even for the 6.5 tubes rupture. In the analysis, the recent version of RELAP5 code was assessed with the test data. It indicated that the calculations agreed well with the measured data and that the plant responses such as the water level and relief valve cycling in the damaged steam generator were reasonably predicted. Finally, sensitivity study on the number of ruptured tubes up to 10 tubes was performed to investigate the coolant release into atmosphere. It indicated that the integrated steam mass released was not significantly varied with the number of ruptured tubes although the damaged steam generator was overfilled for more than 3 tubes rupture. These findings are expected to provide useful information in understanding and evaluating the plant ability to mitigate the consequence of multiple SGTR event.

• Nuclear Engineering and Technology
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• 제28권1호
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• pp.27-35
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• 1996

본 논문에서는, 가압경수로 발전소의 가압기 밀림관내 비정상상태의 열성층현상에 대한 계산 모델을 제안하여 배관내의 온도분포, 유동특성 및 열응력에 대해 연구하였다. 경계면이 시간에 따라 변화가 없거나 정상상태에서 개발된 다른 모델과는 달리 본 모델에서는 고온 및 저온유체 사이의 경계면을 시간의 함수로 가정하였다. 열성층현상에 대한 무차원지배방정식은 SIMPLE 알고리즘을 사용하여 해를 구하였다. 본 수치계산의 결과는 주어진 조건하에서 무차원시간이 약 27.0 일 때 배관의 고온부 및 저온부사이의 최대무차원온도차는 0.78정도이었고, 이때의 열성층 현상에 의한 최대 열응력은 276 MPa로 계산되었다.

• Nuclear Engineering and Technology
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• 제17권2호
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• pp.83-97
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• 1985

GLR (Generalized Likelihood Ratio)방식은 최적상태 변수 추정기인 Kalman-Buchy 필터로부터 발생되는 연속 Innovation들에 대해 통계확률적검사를 수행함으로써 시스템 고장 탐지 및 종류를 판별하게 된다. 그러나, 이러한 종전의 GLR방식은 각 경우마다 특별한 고장 형태를 가정해야 하므로, 모든 가능한 고장 형태를 탐지하는 데 많은 어려움이 있다. 이번 논문에서는 이런 난제를 해결할 한 방법을 제시하였다. 그리고, 가압경수형 원자력발전소 일차측 압력을 조절하는 가압기에 적용시켜 본 결과, 어떤 형태의 고장이든 잘 탐지되고 그 종류도 구별할 수 있음을 보여주었으며, 종전방식에 비해 고장 탐지 및 고장 구별에 필요한 컴퓨터처리 시간도 줄일 수가 있었다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.404-410
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• 1998

In Yonggwang nuclear units 3 and 4 currently under commercial operation, the cold leg temperature was very close to the technical specification limit of 298$^{\circ}C$ during initial startup testing, which was caused by the higher-than-expected reactor coolant system flow. Accordingly, the reference temperature (Tref) program needed to be revised to allow more flexibility for plant operations. In this study, the method of a specific test performed at Yonggwang nuclear unit 4 to revise the Tref program was described and the test results were discussed. In addition, the modified Tref program was evaluated on its potential impacts on system performance and safety. The methods of changing the Tref program and the associated pressurizer level setpoint program were also explained. Finally, for Ulchin nuclear unit 3 and 4 currently under initial startup testing, the effects of reactor coolant system flow rate on the coolant temperature were evaluated from the thermal hydraulic standpoint and an optimum Tref program was recommended.