• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.817-824
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• 2012

The most important component of a nuclear power plant is its nuclear reactor. Studies on the integrity of reactors have become an important part regarding the safety of a nuclear power plant. The US Nuclear Regulatory Commission Regulatory Guide (NRC RG) 1.20 presents a Comprehensive Vibration Assessment Program (CVAP) to be used to verify the structural integrity of the Reactor Vessel Internals (RVI) for flow-induced vibration prior to commercial operation. However, there are few published studies related to the RVI CVAP. We classified the Advanced Power Reactor 1400 (APR1400) RVI CVAP as a non-prototype category-2 reactor as part of an independent validation of its design. The aim of this paper is to present the results of structural response analyses of the Upper Guide Structure (UGS) assembly of the APR1400 reactor. These results show that the UGS and the Inner Barrel Assembly (IBA) meet the specified integrity levels of the design acceptance criteria. The vibration and stress analysis results in this paper will be used as basic information to select measurement locations of the vibration and stress for the APR1400 RVI CVAP.

• Journal of Energy Engineering
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• v.26 no.2
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• pp.12-22
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• 2017

Korea Hydro & Nuclear Power Co. (KHNP) has implemented engineering programs for operating nuclear power plants. Engineering programs are maintenance rule (MR), functional importance determination (FID), single point vulnerability (SPV) and functional equipment group (FEG). Recently, KHNP has developed engineering programs for APR1400 feedwater supplying system to establish the advanced engineering system and will verify the suitability of engineering programs through implementing in new nuclear power plant. Consequently, it is expected that the reliability of APR1400 feedwater supplying system will be improved by implementing engineering programs.

• Proceedings of the KSME Conference
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• 2001.11b
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• pp.505-510
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• 2001

Advanced Pressurized Reactor 1400(APR1400), which is a standard evolutionary advanced light water reactor(ALWR), has been developed from 1992 as one of long-term Government Project(G-7). The APR-1400 is designed to operate at the rated output of 4000MWt to produce an electric power output of around 1450MWe. Due to the increased electric power, In Nuclear Power plant huge quantities of heat are generated in the thermo-dynamic process used for producing electrical energy. So, There is considerationly additional cooling, Heat transfer area and increased cooling water of Heat Exchanger which take care of the different smaller cooling duties within the nuclear power plant. We review applying to PRE instead of Shell-and-Tube Heat exchanger. In this paper, we describe the major design features of PRE, Comparison between a PHE and a Shell-and-Tube Heat Exchanger, and then Applicability of Plate Heat Exchanger in Nuclear Power Plant Component Cooling water systems.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.21 no.1
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• pp.47-55
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• 2011

Reactor vessel internals comprehensive vibration assessment program(RVI CVAP) is one of the necessary tests to ensure the safety of nuclear power plants. RVI CVAP of U.S. nuclear regulatory commission regulatory guide 1.20(U.S. NRC R.G. 1.20) consists of the analysis, measurement and inspection. One of the core technologies of the measurement program for RVI CVAP is to select suitable sensors because the measurement is conducted during the critical path of the construction period of nuclear power plants. Therefore, we analyzed RVI thermal-hydraulic and structure design data of Palo Verde nuclear power plant(U.S.), Yonggwang nuclear power plant(Korea) and APR1400 and researched measuring sensors used in them; moreover, we investigated sensors used for measurement of RVI CVAP for the last 20 years throughout the world. Based on these results, we selected suitable measuring sensors for RVI CVAP in advanced power reactor 1400(APR1400).

• Journal of the Korean Society of Systems Engineering
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• v.19 no.2
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• pp.32-45
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• 2023

Implementing Severe Accident Management (SAM) strategies is crucial for enhancing a nuclear power plant's resilience and safety against severe accidents conditions represented in the analysis of Station Blackout (SBO) event. Among these critical approaches, the In-Vessel Retention (IVR) through External Reactor Vessel Cooling (IVR-ERVC) strategy plays a key role in preventing vessel failure. This work is designed to evaluate the efficacy of the IVR strategy for a high-power density reactor APR1400. The APR1400's plant is represented and simulated under steady-state and transient conditions for a station blackout (SBO) accident scenario using the computer code, ASYST. The APR1400's thermal-hydraulic response is analyzed to assess its performance as it progresses toward a severe accident scenario during an extended SBO. The effectiveness of emergency operating procedures (EOPs) and severe accident management guidelines (SAMGs) are systematically examined to assess their ability to mitigate the accident. A group of associated key phenomena selected based on Phenomenon Identification and Ranking Tables (PIRT) and uncertain parameters are identified accordingly and then propagated within DAKOTA Uncertainty Quantification (UQ) framework until a statistically representative sample is obtained and hence determine the uncertainty bands of key system parameters. The Systems Engineering methodology is applied to direct the progression of work, ensuring systematic and efficient execution.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2014.10a
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• pp.102-105
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• 2014

In accordance with U.S. Nuclear Regulatory Commission regulatory guide(NRC RG) 1.20(Rev.3), we are writing a comprehensive vibration assessment program(CVAP) design control document(DCD) and a technical report for U.S. NRC design certification(DC) of an Advanced Power Reactor 1400(APR1400) nuclear power plant(NPP). CVAP of an APR1400 NPP for U.S. NRC DC is classified as a non-prototype category 1 type. Therefore, CVAP DCD of reactor vessel internals(RVI) and steam generator internals(SGI) consist of analysis and full inspection program. However, piping system of primary and secondary system will be described as measurement program.

• Nuclear Engineering and Technology
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• v.36 no.5
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• pp.388-402
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• 2004

A phenomena identification and ranking table(PIRT) was developed for a main steam line break (MSLB) event for the Advanced Power Reactor-1400 (APR-1400). The selectee event was a double-ended steam line break at full power, with the reactor coolant pump running. The developmental panel selected the fuel performance as the primary safety criterion during the ranking process. The plant design data, the results of the APR-1400 safety analysis, and the results of an additional best-estimate analysis by the MARS computer code were used in the development of the PIRT. The period of the transient was composed of three phases: pre-trip, rapid cool-down, and safety injection. Based on the relative importance to the primary evaluation criterion, the ranking of each system, component, and phenomenon/process was performed for each time phase. Finally, the knowledge-level for each important process for certain components was ranked in terms of existing knowledge. The PIRT can be used as a guide for planning cost-effective experimental programs and for code development efforts, especially for the quantification of those processes and/or phenomena that are highly important, but not well understood.

• Journal of Energy Engineering
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• v.21 no.3
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• pp.292-300
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• 2012

KHNP CRI has been developing APR+ nuclear power plant since 2007, which is GEN III+ model with 4,361 MWth capacity. To develop safer and more economical nuclear power plant than APR1400, we studied domestic and foreign nuclear power plants under construction. We also reviewed nuclear power plants which are appropriate for domestic construction in Korea and also for export. Economic assessments were made twice during the second phase of standard detailed design of the plant. The result of the second phase of economic analysis for APR+ standard detailed design showed that APR+ N-th plant was 24.6% more economical than coal-fired 1,000MW power plant, and was evaluated to be competitive enough in global market for construction of the nuclear power plant.

• Journal of the Korean Society of Systems Engineering
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• v.16 no.2
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• pp.1-15
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• 2020

Nuclear power plants in South Korea are operated to cover the baseload demand. Hence they are operated at 100% rated power and do not deploy power tracking control except for startup, shutdown, or during transients. However, as the contribution of renewable energy in the energy mix increases, load follow operation may be needed to cover the imbalance between consumption and production due to the intermittent nature of electricity produced from the conversion of wind or solar energy. Load follow operation may be quite challenging since the operators need to control the axial power distribution and core reactivity while simultaneously conducting the power maneuvering. In this paper, a systems engineering approach for multi-physics load follow simulation of APR1400 is performed. RELAP5/SCDAPSIM/MOD3.4/3DKIN multi-physics package is selected to simulate the Korean Advanced Power Reactor, APR1400, under load follow operation to reflect the impact of feedback signals on the system safety parameters. Furthermore, the systems engineering approach is adopted to identify the requirements, functions, and physical architecture to provide a set of verification and validation activities that guide this project development by linking each requirement to a validation or verification test with predefined success criteria.

• Journal of Energy Engineering
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• v.23 no.2
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• pp.191-198
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• 2014

Korea Hydro & Nuclear Power Co. (KHNP) has developed and implemented the maintenance effectiveness monitoring (MR) programs for the operating nuclear power plants. MR programs are developed by reflecting design characteristics of the operating nuclear power plants to monitor the plant performance for improving the safety and reliability. Recently, KHNP has developed the MR program for APR1400 safety related systems to establish the advanced maintenance system and will verify the suitability of the MR program through evaluating initial performance. Consequently, it is expected that the safety of the new plant will be improved by developing and implementing the MR program.