• Proceedings of the KSME Conference
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• 2008.11a
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• pp.287-292
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• 2008

The applicability of stainless steel type 316N to the PZR surge-lines of OPR1000 and APR1400 is investigated. So far, strainless steel type 347 has been used for the OPR1000 surge-lines. The degree of improvement in the leak-before-break(LBB) and component design margin is evaluated when stainless steel type 347 is substituted by type 316N. For the study, the tensile and J-R tests on type 316N and type 347 stainless steels were performed at 316 and the microstructure of both types was examined. Stainless steel type 316N shows the higher values on the stress-strain curves, J-R curves and stress intensity, Sm, compared to those of type 347. Therefore, stainless steel type 316N ensures the higher LBB and component design margins. As a result, this study shows that stainless steel type 316N could substitute type 347 for the surge-lines of OPR1000 and APR1400.

• 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 the Korean Society of Safety
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• v.31 no.6
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• pp.129-134
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• 2016

Standard design of APR+(advanced power reactor plus) was certified at 2014 by Korea regulatory body. Based on the experience gained from OPR1000 and APR1400, the APR1400 was being developed as a 1,500MWe class reactor using Korean technologies for design code, reactor coolant pump, and man-machine interface system. APR+ has been basically designed to have the seismic design basis of safe shutdown earthquake (SSE) 0.3g, a 4-train safety concept based on N+2 design philosophy, and a passive auxiliary feedwater system (PAFS). Also, safety issues on the Fukushima-type accidents have been extensively reviewed and applied to enhance APR+ safety. APR+ provides higher reliability and safety against tsunami and earthquake. The purpose of this paper is to implement probabilistic safety assessment considering these design features and to analyze sensitivity of core damage frequency for large loss of coolant accident of APR+.

• Journal of computational fluids engineering
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• v.10 no.3 s.30
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• pp.9-18
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• 2005

During a hypothetical severe accident in a nuclear power plant (NPP), hydrogen is generated by the active reaction of fuel-cladding and steam in the reactor pressure vessel and released with steam into the containment. In order to mitigate hydrogen hazards possibly occurred in the NPP containment, hydrogen mitigation system (HMS) is usually adopted. The design of the next generation NPP (APR1400) designed in Korea specifies 26 passive autocatalytic recombiners and 10 igniters installed in the containment for the hydrogen mitigation. in this study, the analysis of the hydrogen and steam behavior during a total lose of feed water (TLOFW) accident in the APR1400 containment has been conducted by using the CFD code GASFLOW. During the accident, a huge amount of hot water, steam, and hydrogen is released in the in-containment refueling water storage tank (IRWST). The current design of the APR1400 includes flap-type dampers at the IRWST vents which are operated depending on the pressure difference between inside and outside of the IRWST. it was found that the flaps strongly affects the flow structure of the steam and hydrogen in the containment. The possibilities of a flame acceleration and transition from deflagration to detonation (DDT) were evaluated by using Sigma-Lambda criteria. Numerical results indicate the DDT possibility could be heavily reduced in the IRWST compartment when the flaps are installed.

• Nuclear Engineering and Technology
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• v.43 no.3
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• pp.257-270
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• 2011

A set of reflood tests has been performed using ATLAS, which is a thermal-hydraulic integral effect test facility for the pressurized water reactors of APR1400 and OPR1000. Several important phenomena were observed during the ATLAS LBLOCA reflood tests, including core quenching, down-comer boiling, ECC bypass, and steam binding. The present paper discusses those four topics based on the LB-CL-11 test, which is a best-estimate simulation of the LBLOCA reflood phase for APR1400 using ATLAS. Both homogeneous bottom quenching and inhomogeneous top quenching were observed for a uniform radial power profile during the LB-CL-11 test. From the observation of the down-comer boiling phenomena during the LB-CL-11 test, it was found that the measured void fraction in the lower down-comer region was relatively smaller than that estimated from the RELAP5 code, which predicted an unrealistically higher void generation and magnified the downcomer boiling effect for APR1400. The direct ECC bypass was the dominant ECC bypass mechanism throughout the test even though sweep-out occurred during the earlier period. The ECC bypass fractions were between 0.2 and 0.6 during the later test period. The steam binding phenomena was observed, and its effect on the collapsed water levels of the core and down-comer was discussed.

• 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.

• Journal of the Korean Society of Systems Engineering
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• v.12 no.1
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• pp.41-58
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• 2016

Reliability Centered Maintenance (RCM) introduces a systematic method and decision logic tree for utilizing previous operating experience focused on reliability and optimization of maintenance activities. In this paper RCM methodology is applied on safety injection system for APR-1400. Functional Failure Mode Effects and Criticality Analysis (FME&CA) are applied to evaluate the failure modes and the effect on the component, system and plant. Logic Tree Analysis (LTA) is used to determine the optimum maintenance tasks. The results show that increasing the condition based maintenance will reduce component failure and improve reliability and availability of the system. Also the extension of the surveillance test interval of Safety Injection Pumps (SIPs) would lead to an improved pump's availability, eliminate the unnecessary maintenance tasks and this will optimize maintenance activities.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2012.10a
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• pp.300-305
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• 2012

The vibration and stress analysis program of comprehensive vibration assessment program (CVAP) is to verify theoretically the structural integrity of reactor vessel internals (RVI) and to provide the basis for selecting the locations monitored in measurement and inspection programs. This paper covers the verification of the vibration and stress analysis methodology of APR1400 RVI CVAP. The analysis methodology was developed to use 3-dimensional hydraulic and structural models with ANSYS and CFX. To validate the methodology, the hydraulic loads and structural reponses of OPR1000 were predicted and compared with the calculated and measured data in the OPR1000 RVI CVAP. Since the results predicted with this methodology were close to the measured values considerably, it was confirmed that the analysis methodology was developed properly.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.309-316
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• 2024

The U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.20 provides guidance on the comprehensive vibration assessment program (CVAP) to be performed on reactor internals during preoperational and startup tests. The purpose of the program is to identify loads that could cause vibration in the reactor internals and to ensure that these vibrations do not affect their structural integrity. The structural vibrational analysis program involves creating finite element analysis models of the reactor internals and calculating their structural responses when subjected to vibration loads. The appropriateness of the structural analysis methodology must be demonstrated through benchmarks or any other reasonable means. Although existing structural analysis methodologies have been proven to be appropriate and are widely used, this paper presents the development of an improved new structural analysis methodology for APR1400 reactor internals using scaled model tests.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.172-172
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• 2002