• 한국압력기기공학회 논문집
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• 제11권1호
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• pp.74-78
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• 2015

Nuclear power plant steam generator that is one of the main component has several thousands of thin tubes. And the steam generator tube is subject to damage because of the severe operation conditions such as the high temperature and pressure. Therefore periodic inspections are conducted to ensure the integrity of steam generator component. Hanul unit 3 also has been inspected in accordance with in-service inspection program and is scheduled to be replaced for exceeding the plugging rate which was recommended by manufacturer. During the steam generator replacement activity, we found several clustered porosity on inner surface of main feed water pipe. Additionally crack-like indications were found at weld interface between base material and weld of main feed water pipe. This paper describes the field experience and visual testing results for inner surface of main feed water pipes. The destructive test result had shown that these indications were porosities which were caused by manufacturing process not by operation service.

• 대한기계학회논문집A
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• 제37권6호
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• pp.739-745
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• 2013

본 연구에서는 급수 순환 시스템 중 동작 유체의 이송을 위해 사용하는 배관을 대상으로 관로 크기 및 고온 급수 환경에 따른 침식-부식 거동과 그 영향을 규명하고자 하였다. 이를 위해 일반적인 관로 소재인 철과 핵심 부식 인자인 산소의 화학반응식을 기반으로 Hayduk 과 Minhas 가 제안한 모델을 이용하여 침식-부식 해석을 실시하였다. 상용 유한요소해석 프로그램인 ABAQUS 를 사용하여 해석을 수행하였으며 배관의 직경 및 급수 온도를 변화시킴에 따른 침식-부식률을 평가할 수 있었다. 결과를 통해 급수 온도가 침식-부식률에 가장 큰 영향을 미치는 요인이 됨을 알 수 있었으며, 특히 $290^{\circ}C$ 급수에 노출된 스테인리스 316 강은 연간 $2.59{\mu}m$의 두께 손실이 발생할 것으로 예상되었다.

• Journal of Advanced Marine Engineering and Technology
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• 제32권8호
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• pp.1303-1308
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• 2008

Keeping cooling water temperature higher within the allowable range helps marine engines to run in more efficient condition especially when the engine load is low. Temperature control of jacket cooling water in outlet side of main engine has been more widely adopted to ships these days for the purpose to reduce fuel consumption rate. But If the temperature sensor for the control loop is placed at the outlet of engine, it brings more difficulties in attaining stable and desirable properties due to dead times included in pipe length and engine itself comparing to the case where the measuring point is at the inlet side of main engine. In relation with this problem, Feed-forward control could be one of realistic solutions as it reveals good properties and requires less cost for system configuration. This study suggests a forward control system which leads to improved temperature control performances to disturbance signals which could arise from variation of engine load or weather condition. Two dead times in the modelling were described, considering pipe length between the actuator and the engine as well as the thermal process inside the engine. The results of analysis were shown by simulations to confirm responses under different conditions.

• Nuclear Engineering and Technology
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• 제31권4호
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• pp.375-384
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• 1999

The credibility of CHECWORKS FAC model analysis was evaluated for plant application in a model plant chosen for demonstration. The operation condition at each pipe component was defined before the wear rate analysis by plant data base, water chemistry analysis, and network flow analysis. The predicted wear was compared with the measured wear for 57 sample components selected from 43 susceptible line groups analysed. The inspected 57 locations represent components of highest predicted wear in each line group. Both absolute value and relative ranking comparisons indicated reasonable correlations between the predicted and the measured values. Four components showed much higher measured wear rates than the predicted ones in the feed water train from main feed water pump discharge to steam generator, probably due to high hydrazine concentration operation the effect of which had not been incorporated into the CHECWORKS model. The measured wear was higher than the predicted one consistently for components with least susceptibility to FAC. It is believed that the conservatism maintained during UT data analysis dominated the measurement accuracy. A great deal of enhancement is anticipated over the current plant pipe management program when a comprehensive plant pipe management program is implemented based on the model analysis.

• 한국압력기기공학회 논문집
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• 제7권3호
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• pp.40-47
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• 2011

KSNPs(Korea Standard Nuclear Power Plant) have been applied the break exclusion criteria to the high energy lines passing through containment penetration area to ensure that piping failures would not cause the loss of containment isolation function, and to reduce the resulting dynamic effects. Systems with the criteria are the Main Steam system, Feed Water system, Steam Generator Blowdown system, and Chemical & Volume Control system. In accordance with FSAR(Final Safety Analysis Report), a 100% volumetric examination by augmented in-service inspection of all pipe welds appled the break exclusion criteria is required for the break exclusion application piping. However, it is difficult to fully satisfy the requirements of inspection because 12", 8" and 6" weldolet weldments of Main Steam pipe line have complex structural shapes. To resolve the difficulty on the application of conventional UT(Ultrasonic Testing) technique, realistic mock-ups and UT calibration blocks were made. Simulations of conventional UT were performed utilizing CIVA, a commercial NDE(Nondestructive Examination) simulation software. Phased array UT experiments were performed through mock-up including artificial notch type flaws. A phased array UT technique is finally developed to improve the reliability of ultrasonic test at main steam line pipe to 12", 8" and 6" branch connection weld.