• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2005.11a
• /
• pp.475-478
• /
• 2005

Calculation of wear depth with regard to the wear topology is peformed numerically Four typical wear topology, that is round, crescent, flat, and diamond types are adopted to represent the configuration of wear volume. Diamond and flat types are the most severe topology for wear depth history, whereas round and crescent types have small increasing rate of wear depth to the wear volume. Based on this study we can guess that the most severe wear phenomena happens to be upper side of U-tubes in the KSNP SG, because flat or diamond wear will be generated by the wearing motion between tubes and diagonal, vertical, horizontal strips. The misalignment of tube at the stage of manufacturing or distortion of upper structure due to the thermal expansion or vibration of upper structure such as diagonal, vertical, and horizontal strips will be one of the main causes of flat or diamond wear.

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

• Journal of Energy Engineering
• /
• v.13 no.3
• /
• pp.173-180
• /
• 2004

A technique of the heat rate allocation was devised to monitor the performance of Combined Cycle Power Plant. This calculates the expected heat rate of current conditions and compares it with actual values. Loss allocation in heat rate is reconciled by calculating the magnitude of the deficiency contributed by major components, such as the gas turbine, heat recovery steam generator (HRSG), steam turbine and condenser. Expected power output is determined by a detailed model and correction curves of the plant.

• Proceedings of the Korean Institute of Surface Engineering Conference
• /
• 2009.10a
• /
• pp.153-154
• /
• 2009

원자력발전소 증기발생기 전열관 보수 기술의 하나로 니켈 합금 전기 도금이 연구되고 있다. 여러 도금 공정변수 중 peak current density를 달리하여 Ni-P-Fe 전기도금층을 제조한 뒤, 열처리 온도 $325^{\circ}C$에서 10, 30일간 열처리를 한 후, 인장강도와 연신율을 측정하고, 그 파단면을 관찰하였다. 50mA/$cm^2$로 제조된 도금층은 100mA/$cm^2$로 제조된 도금층에 비해 우수한 열적안정성을 가짐을 알 수 있었다.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.05a
• /
• pp.315-320
• /
• 1996

영광 3호기의 정지냉각펌프 성능감시 설비로는 펌프 유량계, 입구압, 출구압, 모터전류 등이 있으며 현장에서 펌프의 소음 감시나 진동 측정 등을 통하여 펌프 건전성을 확인할 수 있다. 부분충수운전중 여러 연구결과 제시된 펌프의 이상징후 증상은 펌프의 소음 증가, 유량계 또는 모터전류의 불규칙 요동이 있으나 정량적인 값을 제시하지 못하고 있으며 공기유입량에 대한 운전제한 근거만 정량적으로 제시되고 있다. 즉, WCAP-l1916에 따른 펌프의 손상 판단 근거는 연속적인 공기 흡입의 경우 2%이내, 간헐적인 공기흡입의 경우 5%를 제시하고 있다. 영광 3 호기의 부분충수운전시 펌프 입구압력을 제외한 다른 펌프 성능감시 변수들은 허용오차 이내로 별다른 펌프 이상 징후를 발견하지 못하였다. 그러나 펌프 입구압력 기록계의 입구압력 및 진동폭 변화는 정지냉각유량률, RCS 수위, 증기발생기 노즐댐 설치 유무에 따라 민감한 변화를 보여주었으며, 펌프의 건전성 감시에 가장 효과적인 변수임을 보여주었다.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.11b
• /
• pp.469-474
• /
• 1996

Alloy 690은 응력부식 균열에 대한 저항성이 요구되는 원자력 발전소 증기발생기 전열관 재료에 사용되고 있다$^{(1)}$ . 응력부식 균열에 대한 저항성은 주로 결정입계에 존재하는 크롬탄화물의 기여에 의한 것이 대부분이다. 크롬탄화 석출물의 핵생성을 알아보기 위해서 110$0^{\circ}C$에서 용체화처리를 0, 1, 3, 10분 동안 하여 관찰하였다. 용체화처리한 모든 시편에서 결정입계에 존재하는 석출물의 분포는 쌍정과 교차하면서 갑자기 변화하는 것을 관찰 할 수 있다. 이처럼 석출물이 존재하지 않는 결정입계들은 대부분 낮은 ∑ 값의 CLS으로부터 약간 벗어난 입계가 될 것이다. 결정입계에 존재하는 석출물은 기지와 Cube-Cube orientation relationship을 갖는다. 그리고 단지 하나의 결정입과 반정합을 이룬다. 기지와 반정합을 이루는 석출물은 M$_{23}$ C$_{6}$형태의 크롬 탄화물이고 격자상수는 기지의 격자상수보다 3배 크다.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.16 no.2
• /
• pp.10-17
• /
• 2020

Securing the integrity of steam generator tubes is an essential requirement for safe operation of nuclear power plants. Therefore, tubes that do not satisfy integrity requirements are no longer usable and must be repaired according to the related requirements. In general, the repair criterion is that the damage depth is more than 40% of the tube wall thickness. However, the plugging limit can be changed and be applied, provided a technical proof is given that integrity can be secured against specific degradation at a specific plants and that approval can be obtained from a regulatory agency. A typical example is alternative repair criteria for defects within the tube sheet or tube support plates. In this paper, a background of establishing the plugging limit for steam generator tubes and changes in maintenance criteria are reviewed as examples.

• Journal of dental hygiene science
• /
• v.2 no.2
• /
• pp.105-113
• /
• 2002

This study was aimed to investigate the practice on the infection control and dental waste disposal in dental clinic. The data used in this study had been collected in 100 dental clinics located in Seoul city from August to October, 2001. It was shown that examination instruments, oral surgery instruments, and periodontal instruments were almost sterilization in dental clinic. It was high in the recognition of individual protection but it was low in the degree of practice. Therefore, the practice in infection control is very important to dental team for safety of patients and operators.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.41 no.3
• /
• pp.171-181
• /
• 2017

Two-phase cross flow exists in many shell-and-tube heat exchangers such as condensers, evaporators, and nuclear steam generators. The drag force acting on a tube bundle subjected to air/water flow is evaluated experimentally. The cylinders subjected to two-phase flow are arranged in a normal square array. The ratio of pitch to diameter is 1.35, and the diameter of the cylinder is 18 mm. The drag force along the flow direction on the tube bundles is measured to calculate the drag coefficient and the two-phase damping ratio. The two-phase damping ratios, given by the analytical model for a homogeneous two-phase flow, are compared with experimental results. The correlation factor between the frictional pressure drop and the hydraulic drag coefficient is determined from the experimental results. The factor is used to calculate the drag force analytically. It is found that with an increase in the mass flux, the drag force, and the drag coefficients are close to the results given by the homogeneous model. The result shows that the damping ratio can be calculated using the homogeneous model for bubbly flow of sufficiently large mass flux.

• Journal of the Korean Society for Nondestructive Testing
• /
• v.30 no.4
• /
• pp.330-337
• /
• 2010

In this paper, 3-dimensional electromagnetic numerical analysis is performed about the eddy current(EC) array probe characteristic which is the next generation probe for accurate diagnosis of steam generator(SG) in nuclear power plants(NPPs). ASME(American Society of Mechanical Engineers) Standard and X-probe combo calibration standard tube are selected for acquisition of eddy current testing(ECT) signals and this result of compared with the real test signals for reasonability of result. Based on the analysis result of calibration standard tube, ECT signals that are about the defects of pitting, stress corrosion cracking(SCC), multiple SCC and wear is obtained. Material of specimen was Inconel 600 which is usually used for SG tubes in NPPs. The operation frequency of 300 kHz were used. The signal characteristics could be observed according to the various defects. The results in this paper can be helpful when the ECT signals from EC array probe are evaluated and analyzed.