• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.579-585
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• 1996

A CATHENA model for the reactor regulating system is developed and tested independently. A CATHENA plant model is created by combining this model with the reference CATHENA model which has been developed to analyze a loss-of-coolant accident (LOCA) for the Wolsong 2 generating station. This model is intended to provide a trip coverage analysis capability. The CATHENA reactor regulating system model includes the demand power routine. the light water zone control absorbers, mechanical control absorbers and adjusters. The CATHENA model is tested for steady state at 103% full power. A postulated accident transient (small LOCA) was also tested. The results show that the control routines in CATHENA were set up properly.

• Nuclear Engineering and Technology
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• v.28 no.3
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• pp.265-273
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• 1996

Among those tests performed during the Yonggwang Nuclear Power Plant Units 3 and 4 (YGN 3&4) Power Ascension Test period, the Loss of a Main Feedwater Pump test at l00% power is one of the major test which characterize the capability of YGN 3&4. In this event, one of the two normally operating main feedwater pumps is tripped resulting in a 50% reduction in the feedwater flow. Unless the NSSS and Turbine/Generator control systems actuate properly, the reactor will be tripped on low SG water level or high pressurizer pressure. The test performed at Unit 3 was successful by meeting all acceptance criteria, and the plant was stabilized at a reduced power level without reactor trip. The measured test data for the major plant parameters are compared with the predictions made by the KISPAC computer code, an updated best-estimate plant performance analysis code, to verify and validate its applicability. The comparison results showed good agreement in the magnitude as well as the trends of the major plant parameters. Therefore, the KISPAC code can be utilized for the best-estimate nuclear power plant design and simulation tool after a further verification using other plant test data.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.641-646
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• 1998

In order to enhance the dynamic and interactive simulation capability of a system thermal hydraulic code for nuclear power plant, applicability of flow network models in SINDA/FLUIN $T^{™}$ has been tested by modeling feedwater system and coupling to DSNP which is one of a system thermal hydraulic simulation code for a pressurized heavy water reactor. The feedwater system is selected since it is one of the most important balance of plant systems with a potential to greatly affect the behavior of nuclear steam supply system. The flow network model of this feedwater system consists of condenser, condensate pumps, low and high pressure heaters, deaerator, feedwater pumps, and control valves. This complicated flow network is modeled and coupled to DSNP and it is tested for several normal and abnormal transient conditions such turbine load maneuvering, turbine trip, and loss of class IV power. The results show reasonable behavior of the coupled code and also gives a good dynamic and interactive simulation capabilities for the several mild transient conditions. It has been found that coupling system thermal hydraulic code with a flow network code is a proper way of upgrading simulation capability of DSNP to mature nuclear plant analyzer (NPA).

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.37-43
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• 2014

Nuclear power plants(NPPs) are consisted of power production functions and safety functions preventing leakage of radiation. Operators working in NPPs shall maintain these functions during an operation period through various activities such as improvement & modification, corrective maintenance, preventive maintenance and surveillance test. According to the performance of these work activities, there are configuration changes in NPPs systems. Its changes cause the increase of safety risks(CDF) and plant trip risks. Recently, the importance of risk management is increasing gradually in the operation process of NPPs. Therefore, this paper presents the work management methods using the various risk monitoring systems during power operation and overhaul period. Also this paper suggests the optimum application ways of risk systems for work management.

• 유체기계공업학회:학술대회논문집
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• 2005.12a
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• pp.607-612
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• 2005

Check valves playa vital role in the operation and protection of nuclear power plants. Check valves failure in nuclear power plants often lead to a plant transient or trip. An overview of the failure history of check valves needs to identify key area where resources can be best applied to further improve their reliability, and provide cost effective means for failure reduction. The analysis of historical failure data gives information on the populations of various types of check valves, the systems they are installed in, failure modes, effects, methods of detection, and the mechanisms of the failures. The results presented are based on information derived from operating records, nuclear industry reports, manufacturer supplied information. A majority of check valve failures are caused by improper application. Failure modes are identified for swing and lift check valves. Failures involving improper seating and valve disc stuck comprised the largest percentage of failures.

• 한국신재생에너지학회:학술대회논문집
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• 2008.05a
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• pp.364-367
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• 2008

This present paper describes concepts of the real-time operation monitoring system as a tool for enhancing the reliability and raising the availability of the first Korea IGCC (Integrated Gasification Combined Cycle) power plant. This system consists of five (5) modules : (1) Data Validation Module, (2) Performance Calculation Module (3) Performance Diagnostic Module, (4) Trip Information Module, and (5) Statistics Analysis Module. Among these modules, Performance Calculation Module is explained in more detail. The objective of this module is to continuously evaluate the degradation (decrease in performance) of the IGCC plant and its equipment in order to provide plant operators additional information to help them identify problems, improve performance.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.10 no.8
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• pp.1824-1831
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• 2009

This paper analyzes the propagation speed of the wide-area power system frequency. When a generator supplying the electric power to the grid is accidently tripped due to a disturbance on the systems, power system frequency suddenly drops during the transient period and this propagates from the location of the tripped generator to the other part of the systems like a wave. Since the propagation speed of the power system frequency depends on the own characteristics of power systems, its understanding from the perspective of the wide-area can help us in understanding power systems more correctly. In addition, the propagation speed of the power system frequency is used as a key parameter in the application study of IT based on the internet-based GPS synchronized frequency monitoring network (FENT) which has been recently implemented and operated in U.S. power systems. This paper simulates the generation trip on various locations in U.S. power systems deploying its latest dynamic model and calculates the propagation speed of the power system frequency for the application of FNET.

• Nuclear Engineering and Technology
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• v.30 no.3
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• pp.222-234
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• 1998

The Westinghouse OT$\Delta$T DNB protection logic heavily restricts the operation region by applying the same logic for a full range of operating pressure in order to maintain its simplicity. In this work, a fuzzy neural network method is used to estimate the DNB protection limit using the measured average temperature and pressure of a reactor core. Fuzzy system parameters are optimized by a hybrid learning method. This algorithm uses a gradient descent algorithm to optimize the antecedent parameters and a least-squares algorithm to solve the consequent parameters. The proposed method is applied to Yonggwang 3&4 nuclear power plants and the proposed method has 5.99 percent larger thermal margin than the conventional OT$\Delta$T trip logic. This simple algorithm provides a good information for the nuclear power plant operation and diagnosis by estimating the DNB protection limit each time step.

• Journal of Korean Society of Occupational and Environmental Hygiene
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• v.26 no.3
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• pp.377-387
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• 2016

Objectives: The objective of this study was to describe the emotional labor experiences of ward nurses in Korea. Methods: The data were collected through focus group interviews with seven ward nurses. The phenomenological analytic method suggested by Colaizzi was used to analyze the data. Results: Four theme clusters and eleven themes emerged from the analysis. The first theme cluster, 'Terrible working environment' includes changes in the health environment and realities of nursing organization. The second theme cluster, 'Suppressed feelings' includes institutionalized power trip, patient-centered emotion management and burning. The third theme cluster, 'Tired body and mind' includes withdrawal, wounded self-esteem, burnout and turnover desicion. The fourth theme cluster, 'Want to be comforted' includes emotional control and wishing to live as a nurse. Conclusions: The findings of this study provide an understanding of the emotional labor experiences of ward nurses to explain their perceptions and language. The essence of the emotional labor of ward nurses is a negative working environment created by increased competition in the healthcare industry and the nature of nursing organization, and as a result their quality of life has fallen.

• Journal of Energy Engineering
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• v.21 no.1
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• pp.75-80
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• 2012

In this paper a realistic bounding method for flooding analysis following rupture of large size of thanks and piping is defined. Mass and energy release during main feedwater line break accident is analyzed with RETRAN code. It is modeled that the main feed water control valve is closed in 5.0 seconds after reactor trip. In result of the analysis, largest mass and energy is discharged at 70% reactor power. The flood sources for main feedwater room are calculated when piping failure occurs in the high energy line and medium energy line. Based on the result of flood level (1.43m), it is investigated that all of the safety-related environmental qualification equipments are well located above the flood level.