• Proceedings of the KIEE Conference
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• 2008.07a
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• pp.1667-1668
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• 2008

There are three main devices such as boiler producing steam, turbine driving generator and generator producing electricity. An electrical generator in power plant is driven and maintained its speed at rated by steam turbine which is coupled into generator directly. Turbine auxiliary devices such as condenser, deaerator, feed water heater, gland steam condenser, pump recirculation equipment, feed water pump, and so on should be operated well so that the steam turbine exert its maximum efficiency. There are many control loop such as hot well level and condenser recirculation, deaerator level, pegging steam pressure, feed water heater level, feed water pump recirculation. In this paper condenser level control and deaerator level control are going to be described.

• Journal of the Korean Society of Manufacturing Process Engineers
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• v.13 no.3
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• pp.42-48
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• 2014

The auxiliary-feed-water pump (AFWP) used to supply water during a station black out situation at nuclear power plants should meet the seismic qualification regulations stipulated in IEEE Std 323 and 344, so as to withstand earthquakes or dangerous situations. Here, we establish a model for the estimation of the structural integrity of this type of pump. If the natural frequency that results from a modal analysis is less than 33 Hz, we adopt a dynamic analysis, instead of a static analysis. A dynamic analysis was carried out taking into consideration seismic conditions such as the floor response spectra (FRS), an operation-base earthquake (OBE), and a safe-shutdown earthquake (SSE). Finally, an analytical estimation of the structural integrity of an AFWP is made through a comparison of calculated values and allowable values. If the result is less than the allowable stress, the pump is deemed to have good structural integrity. In addition, future studies will involve a stability check for rotor accidents that may occur during the operation of the pump.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.514-519
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• 1997

The MRX is an integral type ship reactor with 100 MWt power, which is designed by Japan Atomic Energy Research Institute. It is characterized by integral type PWR, in-vessel type control roe drive mechanism, water-filled containment vessel and passive decay heat removal system. Marine reactor should have high passive safety. Therefore, in this study, we simulated the loss of flow accident to verify the passive decay heat removal by natural circulation using RETRAN-03 code. auxiliary feed water systems are used for decay heat removal mechanism and results are compared with the loss of flow accident analysis using emergency decay heat removal system by JAERI. Results are very similar to case of EDRS 1 loop operation in JAERI analysis and decay heat is successfully removed by natural circulation.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.36 no.5
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• pp.553-562
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• 2012

Computational Fluid Dynamics (CFD) analysis has been performed to estimate the air temperature inside an Auxiliary Feed Water (AFW) Motor-Driven (MD) pump room for the case where there is loss of Heating, Ventilation, and Air-Conditioning (HVAC). A transient calculation for the closed pump room without cooling by any HVAC system shows that the volume-averaged air temperature reaches around $60^{\circ}C$ for a transient period of 8.0 h. From previous studies, the external air and surface boundary temperatures are assumed to increase slowly starting from an initial temperature of $35^{\circ}C$. For the cases where the door is opened at 2, 4, and 6 h after the initiation of HVAC failure, the average air temperature promptly drops by about $4^{\circ}C$ when the door is opened and then slowly increases. The current calculations based on the CFD technique predict the rate of increase of air temperature to be lower than that determined by previous conservative calculations on the basis of a lumped model.

• Journal of the Korean Society of Combustion
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• v.19 no.3
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• pp.8-17
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• 2014

The aims of this research were to evaluate effects of biomass co-firing to pulverized coal power plants and the variation of co-firing ratios on the plant efficiency related to power consumption of auxiliary system and flue gas characteristics such as production and component by process simulation based on the existing pulverized coal power plant. In this study, four kinds of biomass are selected as renewable fuel candidates for co-firing: wood pellet(WP), palm kernel shell(PKS), empty fruit bunch(EFB) and walnut shell(WS). Process simulation for various biomass fuels and co-firing ratios was performed using a commercial software. Gas side including combustion system and flue gas treatment system was considering with combination of water and steam side which contains turbines, condenser, feed water heaters and pumps. As a result, walnut shell might be the most suitable as co-firing fuel among four biomass since when 10% of walnut shell was co-fired with 90% of coal on thermal basis, flue gas production and power consumption of auxiliary systems were the smallest than those of other biomass co-firing while net plant efficiency was relatively higher than those of other biomass co-firing. However, with increasing walnut shell co-firing ratios, boiler efficiency and net plant efficiency were expected to decrease rather than coal combustion without biomass co-firing.

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1603-1618
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• 2024

Emergency operation plans are indispensable elements for effective process safety management especially when unanticipated events occur under extreme situations. In the paper, a synthesis framework is proposed for the integration success path planning and performance analysis. Within the synthesis framework, success path planning is implemented through flow-directed signal tracing, renaming and reconstruction from a complete collection of Minimal Path Sets (MPSs) that are obtained using graph traversal search on GO-FLOW model diagram. The performance of success paths is then evaluated and prioritized according to the task complexity and probability calculation of MPSs for optimum action plans identification. Finally, an Auxiliary Feed Water System of Pressurized Water Reactor (PWR-AFWS) is taken as an example system to demonstrate the flow-directed MPSs search method for success path planning and performance analysis. It is concluded that the synthesis framework is capable of providing procedural guidance for emergency response and safety management with optimal success path planning under extreme situations.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.60-72
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• 2019

The air-cooled passive decay heat removal system (APDHR) was proposed to provide the ultimate heat sink for non-LOCA accidents. The APDHR is a modified one of Passive Auxiliary Feed-water system (PAFS) installed in APR+. The PAFS has a heat exchanger in the Passive Condensate Cooling Tank (PCCT) and can remove decay heat for 8 h. After that, the heat transfer rate through the PAFS drastically decreases because the heat transfer condition changes from water to air. The APDHR with a vertical heat exchanger in PCCT will be able to remove the decay heat by air if it has sufficient natural convection in PCCT. We conducted the thermal-hydraulic simulation by the MARS code to investigate the behavior of the APR + selected as a reference plant for the simulation. The simulation contains two phases based on water depletion: the early phase and the late phase. In the early phase, the volume of water in PCCT was determined to avoid the water depletion in three days after shutdown. In the late phase, when the number of the HXs is greater than 4089 per PCCT, the MARS simulation confirmed the long-term cooling by air is possible under extended Station Blackout (SBO).

• Nuclear Engineering and Technology
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• v.37 no.1
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• pp.109-117
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• 2005

Limiting conditions for operations (LCOs) are evaluated dynamically using the tool of system dynamics. The LCOs de-fine the allowed outage times (AOTs) and the actions to be taken if the repair cannot be completed within the AOT. System dynamics has been developed to analyze the dynamic reliability of a complicated system. System dynamics using Vensim software have been applied to LCOs assessment for an example system, the auxiliary feed water system of a reference nuclear power plant. Analysis results of both full power operation and shutdown operation have been compared for a measure of core damage frequency. The framework developed in this study has been shown to be very flexible in that it can be applied to assess LCOs quantitatively under any operational context of the TS in FSAR.

• Fire Science and Engineering
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• v.19 no.3 s.59
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• pp.20-27
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• 2005

This paper evaluates the fire safety level of eight pump rooms in the nuclear power plant using a fire model, CFAST We estimate the Conditional Core Damage Probability (CCDP) of each room based on the analyzed results of CFAST Eight rooms located on the primary auxiliary building of the nuclear power plant are high pressure safety injection pump room A/B, low pressure safety injection pump room Am. containment sprdy pump room A/B, and motor-driven auxiliary feed water pump room A/B. The upper layer gas temperature of each room is estimated and the integrity of cable is reviewed. Based on the results, the integrity of the cable located at the upper part of compartment is maintained without thermal damage. The Conditional Core Damage Probability Is reduced to half of the old values. Accordingly, the fire safety assessment for eight pump rooms using the fire model will be capable of reducing the uncertainty and to develop a more realistic model.