• Nuclear Engineering and Technology
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• 제49권2호
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• pp.313-326
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• 2017

The systematic management of plant risk is crucial for enhancing the safety of nuclear power plants and for designing new nuclear power plants. Accident sequence precursor (ASP) analysis may be able to provide risk significance of operational experience by using probabilistic risk assessment to evaluate an operational event quantitatively in terms of its impact on core damage. In this study, an ASP methodology for two operation mode, full power and low power/shutdown operation, has been developed and applied to significant accident precursors that may occur during the operation of nuclear power plants. Two operational events, loss of feedwater and steam generator tube rupture, are identified as ASPs. Therefore, the ASP methodology developed in this study may contribute to identifying plant risk significance as well as to enhancing the safety of nuclear power plants by applying this methodology systematically.

• 전기학회논문지
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• 제60권5호
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• pp.1017-1022
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• 2011

This paper describes the improvement of the calculation by using $6{\sigma}$ method on steam turbine simulator in a nuclear power plant. The simulator is essential to not only verification and validation of control logic but also making sure of control constants in upgrading the long time used control system into the new one. And the dynamic model is a key point in that simulator. The model used during the retrofit period of the turbine controller in Kori Nuclear Power Plant makes difference in calculating generator output and control valve positions. That is because such operating data as the main steam pressure, the main steam temperature and control valve positions of Yongkwang #3 are different from those of Kori #4. Therefore, the model parameters must be tuned by using actual operating data for the high fidelity of simulator in calculating the dynamic characteristic of the model. This paper describes that the $6{\sigma}$ method is used in improvement of precision of generator output calculation in the steam turbine model of the simulator.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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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).

• 한국연소학회지
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• 제18권3호
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• pp.44-53
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• 2013

In this study, the economic evaluation for imported coals was conducted for power plant based on thermo-dynamical performance analysis. The number of coal types considered was 1,755 imported by five power generation companies in Korea during the 2010-2012. The higher heating value (HHV) of the coals ranged 4,000-6,500 kcal/kg, mostly sub-bituminous. The 1D thermo-dynamical performance modeling was performed for a 500 MWe standard power plant using PROATES code. It was founded that the low rank coals had negative effects on the plant efficiency mainly due to the increased heat loss by moisture, hydrogen and flue gas. Based on the performance analysis, the economic performance of the coals was evaluated. The apparent price of low-rank coals tended to be significantly lower than design coal; for example, the unit price of coal with a HHV of 4,000 kcal/kg was 57% of the reference coal having 6,080 kcal/kg. Considering the negative effects leading to a decrease in the thermal performance, heating value compensation, and increased parasite load, the corrected unit cost for the coal with 4,000 kcal/kg was 90.7% of the reference coal. Overall, the cost saving by imported coals was not high as expected.

• 전기학회논문지
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• 제60권4호
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• pp.830-837
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• 2011

A turbine control system which has been operated for years in a nuclear power plant was retrofitted with a newly developed digital control system. After completion of the retrofit, turbine valve tests were performed to ensure the integrity of each valve's control function. The sequence of each valve test is composed of a closing process and a reopening process. To minimize megawatt variation which normally occurs during the test sequence, we employed a kind of compensator algorithm in the new digital control system which also have been used in the old system. There were difficulties finding optimal parameter settings for our new compensator algorithm because the power plant didn't allow us to perform necessary tuning procedures while the turbine is on load operation. Therefore an alternative measure for the compensator tuning which is independent of the turbine actual operation had to be implemented. So, a process model for the test was required to overcome this situation. We analyzed the operation data of the test and implemented the process model by use of input and output variable relations. Also we verified the process model by use of another condition's operating data. The result shows that the output of model is similar to the actual operation data.

• 신재생에너지
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• 제5권4호
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• pp.39-43
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• 2009

The variation of inflow at stream and hydrologic performance for small scale hydro power(SSHP) plants due to climate change have been studied. The model, which can predict flow duration characteristic of stream, was developed to analyze the variation of inflow caused from rainfall condition. And another model to predict hydrologic performance for SSHP plants is established. Monthly inflow data measured at Andong dam for 32 years were analyzed. The existing SSHP plant located in upstream of Andong dam was selected and analyzed hydrologic performance characteristics. The predicted results from the developed models show that the data were in good agreement with measured results of long term inflow at Andong dam and the existing SSHP plant. Inflow and ideal hydro power potential had increased greatly in recent years, however, these did not lead annual energy production increment of existing SSHP plant. As a results, it was found that the models represented in this study can be used to predict the primary design specifications and inflow of SSHP plants effectively.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.587-591
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• 1997

The DGs (Diesel Generators) in NPP (Nuclear Power Plant) has been used for the emergency electric power source to shut down the nuclear reactor safely in case of station blackout. The RCM (Reliability Centered Maintenance) has been applied to DGs for increasing the safety of NPP. The structured defects of DG were not remedied by the improvement of maintenance method. As the first stage of RCM, to find the structured defects, its failure modes were searched and analyzed through the ten year maintenance information. The structured defects such as the air compressor, the lubricating oil pressure, and the insufficient load were the root causes of main failures. The air reservoir reinstallation, the lubricating oil tube modification, the load bank installation, and the qualitative instrumentation were the solutions for the hardware oriented RCM of DGs. There remains the software oriented RCM such as the rejection of useless maintenance, the preventive maintenance, the database of maintenance information, and the predictive maintenance.

• 전기학회논문지
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• 제63권12호
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• pp.1637-1639
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• 2014

In the past, the analog protection relays have been widely used for the safety-related systems in the nuclear power plants due to their stability and reliability. Meanwhile, as the high performance digital system has been developed, the digital systems have been adopted in the non-safety systems. However, since the digital systems currently used in the non-safety systems were not developed according to Q-class standard, Commercial Grade Item Dedication (CGID) procedure should be performed in order to apply them to the safety-related system. The purpose of this paper is to describe the CGID procedure including the analysis of the hardware architecture as well as the software embedded in protective relay to apply to the emergency diesel generator in the nuclear power plant. The entire CGID procedure was performed strictly according to the international standard and regulations.

• 전기학회논문지
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• 제64권3호
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• pp.375-383
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• 2015

The virtual power plant (VPP) is a new technology to achieve flexibility as well as controllability, like traditional centralized power plants, by integrating and operating different types of distributed energy resources (DER) with the information communication technology (ICT). Though small-sized DERs may not be controlled in a centralized manner, these are more likely to be utilized as power plants for centralized dispatch and participate in the energy trade given that these are integrated into a unified generation profile and certain technical properties such as dispatch schedules, ramp rates, voltage control, and reserves are explicitly implemented. Unfortunately, the VPP has been in a conceptual stage thus far and its common definition has not yet been established. Such a lack of obvious guidelines for VPP may lead to a further challenge of coming up with the business model and reinforcing the investment and technical support for VPP. In this context, this paper would aim to identify the definition of VPP as a critical factor in smart grid and, at the same time, discuss the details required for VPP to actively take part in the electricity market under the smart grid paradigm.

• Nuclear Engineering and Technology
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• 제55권11호
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• pp.4252-4265
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• 2023

In the case of nuclear power plants near end of their design life, a reassessment of the performance of safetyrelated equipment may be necessary to determine whether to shut down or extend the operation of the power plant. Therefore, it is necessary to evaluate the level of performance decline due to degradation. Electrical cabinets, including MCC and switchgear, are representative safety-related equipment. Several studies have assessed the degradation and seismic performance of nuclear power plant equipment. Most of those researches are limited to individual components due to the size of safety-related equipment and test equipment. However, only a few studies assessed the degradation performance of electrical cabinets. The equipment of various nuclear power plants is anchored to concrete foundations, and crack in concrete foundations is one of the most representative of degradation that could be visually confirmed. However, it is difficult to find a study for analysis through testing the effect of cracks in concrete foundations on the response of electrical cabinet internal equipment fixed by anchors. In this study, using a simple cabinet model considering cast-in-place anchor in uncracked and cracked concretes, a tri-axial shaking table tests were performed and the seismic behavior were observed.