• Nuclear industry
• /
• v.7 no.2 s.48
• /
• pp.41-43
• /
• 1987

Commonwealth Edison사는 미국 굴지의 원자력발전전력회사중 하나로 공인되고 있다. 이와 같이 성공을 거두게 된 이유중에는 2,100만달러의 훈련센터를 포함하는 교육훈련과 품질보증에 혼신의 노력을 경주하고 Inpo, Numarc, EPRI 및 Nsac등과 같은 국가기관에 적극적으로 참여하였기 때문이다. 다음은 이 회사의 사장인 James J. O' Connor씨가 이와 같은 품질개선을 추구한데 대해서 회고한 내용이다.

• Nuclear Engineering and Technology
• /
• v.39 no.2
• /
• pp.133-142
• /
• 2007

The Electric Power Research Institute (EPRI) demonstrated a method for monitoring the performance of instrument channels in Topical Report (TR) 104965, 'On-Line Monitoring of Instrument Channel Performance.' This paper presents the results of several models originally developed by EPRI to monitor three nuclear plant sensor sets: Pressurizer Level, Reactor Protection System (RPS) Loop A, and Reactor Coolant System (RCS) Loop A Steam Generator (SG) Level. The sensor sets investigated include one redundant sensor model and two non-redundant sensor models. Each model employs an Auto-Associative Kernel Regression (AAKR) model architecture to predict correct sensor behavior. Performance of each of the developed models is evaluated using four metrics: accuracy, auto-sensitivity, cross-sensitivity, and newly developed Error Uncertainty Limit Monitoring (EULM) detectability. The uncertainty estimate for each model is also calculated through two methods: analytic formulas and Monte Carlo estimation. The uncertainty estimates are verified by calculating confidence interval coverages to assure that 95% of the measured data fall within the confidence intervals. The model performance evaluation identified the Pressurizer Level model as acceptable for on-line monitoring (OLM) implementation. The other two models, RPS Loop A and RCS Loop A SG Level, highlight two common problems that occur in model development and evaluation, namely faulty data and poor signal selection

• Journal of the Earthquake Engineering Society of Korea
• /
• v.24 no.2
• /
• pp.103-110
• /
• 2020

Seismic qualification of instruments and devices mounted on electrical cabinets in a nuclear power plant is performed in this study by means of the in-cabinet response spectrum (ICRS). A simple method and two rigorous methods are proposed in the EPRI NP-7146-SL guidelines for generating the ICRS. The simple method of EPRI can give unrealistic spectra that are excessively conservative in many cases. In the past, the time domain analysis (TDA) methods have been mostly used to analyze a structure. However, the TDA requires the generation of an artificial earthquake input motion compatible to the target response spectrum. The process of generating an artificial earthquake may involve a great deal of uncertainty. In addition, many time history analyses should be performed to increase the accuracy of the results. This study developed a numerical analysis program for generating the ICRS by frequency domain analysis (FDA) method. The developed program was validated by the numerical study. The ICRS calculated by FDA thoroughly matched with those obtained from TDA. This study then confirms that the method it proposes can simply and efficiently generate the ICRS compared to the time domain method.

• Corrosion Science and Technology
• /
• v.11 no.5
• /
• pp.173-178
• /
• 2012

To manage the wall thinning of carbon steel piping in nuclear power plants, the utility of Korea has performed thickness inspection for some quantity of pipe components during refueling outages and determined whether repair or replacement after evaluating UT (Ultrasonic Test) data. When the existing UT data evaluation methods, such as Band, Blanket, PTP (Point to Point) Methods, are applied to a certain pipe component, unnecessary re-inspecting situations may be generated even though the component does not thinned. In those cases, economical loss caused by repeated inspection and problems of maintaining the pipe integrity followed by decreasing of newly inspected components may be generated. EPRI (Electric Power Research Institute) in USA has suggested several statistical methods, TPM (Total Point Method), LSS (Least Square Slope) Method, etc. to distinguish whether multiple inspecting components have thinned or not. This paper presents the analysis results for multiple inspecting components over three times based on both NAM (Near Area of Minimum) Method developed by KEPCO-E&C and the other methods suggested by EPRI.

• Nuclear Engineering and Technology
• /
• v.31 no.4
• /
• pp.375-384
• /
• 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.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.27 no.2
• /
• pp.239-246
• /
• 2003

This paper proposes engineering estimation equations of elastic-plastic J and COD fur axial through-wall cracked pipes under internal pressure. Based on detailed 3-D FE results based on deformation plasticity, the plastic influence functions for fully plastic J and COD solutions are tabulated as a function of the mean radius-to-thickness ratio, the normalized crack length. and the strain hardening. Based on these results, the GE/EPRI-type J and COD estimation equations are proposed and validated against the 3-D FE results based on deformation plasticity. For more general application to general stress-strain laws or to complex loading, the developed GE/EPRI-type solutions are re-formulated based on the reference stress concept. Such a reformulation provides simpler equations for J and COD, which are then further extended to combined internal pressure and bending. The proposed reference stress based J and COD estimation equations are compared with elastic-plastic 3-D FE results using actual stress-strain data for Type 316 stainless steels. The FE results for both internal pressure cases and combined internal pressure and bending cases compare very well with the proposed J and COD estimations.

• Journal of the Korea Academia-Industrial cooperation Society
• /
• v.14 no.4
• /
• pp.2043-2048
• /
• 2013

OPR-1000 CE Steam Generator, of which tube material is composed of Alloy-600 HTMA in nuclear power plant, secondary side is generated ODSCC(Outside Diameter Stress Corrosion Cracking) due to the accumulated sludge. ODSCC is centered around the tube sheet and is being affected depending on the height of the sludge. Chemical cleaning was carried out for a top-of-the-tube sheet(TTS) of Steam Generator in order to decrease corrosive condition of the secondary side of Steam Generator tubes and suppress the occurrence of stress corrosion cracking. The amount of sludge removal was 259.2kg. The height of the accumulated sludge was reduced from 0.71 to 0.34 inches. Corrosion rate as the maximum 2.34 mils was satisfied to within EPRI (Electric Power Research Institute) recommendation(10 mils).

• 한국IT서비스학회:학술대회논문집
• /
• 2009.11a
• /
• pp.321-326
• /
• 2009

원자력발전 산업은 국내 최대의 전력공급원으로서 '09년 현재 20기의 원전이 가동 중에 있고 8기의 원전이 건설 중이다. 또한 수출주도형 독자노형 개발을 목적으로 한 원전 설계 개발 사업도 진행되고 있다. 국가 기반산업이기도 한 원자력 산업에는 다양한 정보관리시스템들이 운영되고 있으며 지속적인 개선이 이루어지고 있다. 원자력발전소는 60년에 이르는 긴 수명주기를 가지고 있기 때문에 설계 단계에서부터 생산되는 정보를 관리할 수 있는 체계적인 방법이 필요하며, 산업 특성상 다양한 참여기관들이 상호 협업을 진행할 수 있는 통합 창구 역할의 시스템이 요구되고 있다. 본 연구에서는 원자력발전소의 정보관리시스템 개발을 위해 신규원전 사업자에게 요구되는 EPRI URD 요건 및 실제 원전 설계개발 사업에서의 시스템 요구사항 등을 분석하여, 원자력발전소의 원전 정보관리시스템 개발요건을 도출하였으며 ISO 10303 국제표준 기술을 활용한 일부 시스템 프로토타입 구현을 통한 예비 타당성 평가를 수행하였다. 본 연구를 통해 국내 원전 산업에 국제표준 도입 가능성은 확인하였으나, 국제표준 개발이 아직 진행 중인 점을 고려할 때 실제 원전 적용을 위해서는 향후 추가 연구가 필요하다.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.31 no.6 s.261
• /
• pp.718-724
• /
• 2007

Some of gas turbine model of 7F-Class has constructed and is operating with units domestically. Non-destructive testing (NDT) is one of the methods being used to inspect damage $1^{st}$ stage bucket and review damage trends. We also analyze damage configuration and microstructure according to material and compare with pape of electric power research institute (EPRI). The damaged mode could be determined by leveraging failure analysis. Especially, configuration uprate of bucket is not only to prevent damage during operation but also avoid domestic manufacturing by the competitors. Modifications were mainly concentrated on surfaces such as cooling hole and bucket tips. Analyzing of bucket damage, the earlier model of 7F-Class used with one cycle with equivalent operation hour (EOH), has various cracking of the bucket surface. Bucket damage of new model is centered on tip area (54%) as analyzed by EPRI research. We conclude that improving bucket configuration would increase repair rate on the bucket tip.

• 유체기계공업학회:학술대회논문집
• /
• 2003.12a
• /
• pp.549-554
• /
• 2003

EPRI PPM (Performance Prediction Methodology), a method used for the prediction of required thrust of valves, can not be applied to unbalanced-disk globe valves operated in the fluid when the fluid temperature is above $150^{\circ}F$ because the thrust increase caused by the friction between the valve disk and body is not considered in the PPM. In order to apply PPM to the valves, EPRI suggested new friction prediction method to be added in the code. This paper analyzes the applicability of the prediction method comparing the disk-to-body friction load predicted from the method with the measured friction load from the field tests. The maximum values from the prediction method and those obtained from the test were 268lbs and about 1500lbs, respectively. It is included that the prediction method should be improved for the realistic prediction of disk-to-body friction load.