• 시스템엔지니어링학술지
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• 제14권2호
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• pp.58-66
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• 2018

This paper represents a systematic approach aimed at improving the performance of the proportional integral (PI) controller for the Advanced Power Reactor (APR) 1400 Feedwater Control System (FWCS). When the performance of the PI controller offers superior control and enhanced robustness, the steam generator (SG) level is properly controlled. This leads to the safe operation and increased the availability of the nuclear power plant. In this paper, a systems engineering approach is used in order to design a novel PI controller for the FWCS. In the reverse engineering stage, the existing FWCS configuration, especially the characteristics of the feedwater controller as well as the feedwater flow path to each SG from the FWCS, were reviewed and analysed. The overall block diagram of the FWCS and the SG was also developed in the reverse engineering process. In the re-engineering stage, the actual design of the feedwater PI controller was carried out using a genetic algorithm (GA). Lastly, in the validation and verification phase, the existing PI controller and the PI controller designed using GA method were simulated in Simulink/Matlab. From the simulation results, the GA-PI controller was found to exhibit greater stability than the current controller of the FWCS.

• 에너지공학
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• 제23권4호
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• pp.236-239
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• 2014

한수원 중앙연구원은 2007년부터 정부과제의 하나로 전기출력이 1500MWe급인 GEN.III+ 원전 APR+를 개발하고 있다. APR1400에 비해 보다 개선된 안전성과 경제성을 갖는 원전을 개발하기 위해 국내 건설 중인 원전과 해외에서 개발 또는 건설 중인 ALWR(Advanced Light Water Reactor)의 설계내용 및 후쿠시마 원전사고로부터 도출된 개선사항을 반영하여 한국 실정에도 맞고, 해외 수출형 원전에도 부합되는 원전을 설계하고 있다. APR+의 경쟁력을 확인하기 위해 APR+ 표준설계개발 단계에서 3회의 경제성 평가를 수행하였다. 표준설계개발 단계에의 3차(최종) 경제성 평가 결과 APR+ N-th호기는 국내석탄화력 1000MWe급 대비 약 23% 경쟁력 우위인 것으로 평가되었다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2007년도 심포지엄 논문집 정보 및 제어부문
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• pp.305-306
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• 2007

본 논문은 차세대 원전 신고리 #3, 4호기 NSSS(Nuclear Steam Supply System) 제어계통의 검증시스템을 개발하기 위한 설계개념에 관한 것이다. 차세대 원전 신고리 #3, 4호기는 KHNP(Korea Hydro & Nuclear Power Co., Ltd.)가 개발한 APR1400(Advanced Power Reactor 1400 [MWe])을 적용하는 최초의 원자력 발전소이다. APR1400은 3세대 원자력발전소로 인정받고 있으며, APR1400 원자력발전소의 안전한 운영을 위하여 I&C(Instrumentation and Control)시스템이 디지털 표준 플랫폼으로 설계되었다[2]. 특히, 차세대 원전 신고리 #3, 4호기의 비안전계통(제어 감시 및 경보계통)은 WEC (Westinghouse Electric Company)의 DCS(Distributed Control System) 상용 단일 플랫폼으로 구성될 예정이다. 우리는 신고리 #3, 4호기의 제어계통 중에서 NSSS(Nuclear Steam Supply System) 제어계통의 검증시스템을 개발하기 위하여 Stimulated Simulator의 방법론을 적용하여 "Simulator"라는 설계 개념을 정립하였다. 현재 원자력발전소 NSSS 제어계통의 DCS Stimulator 개발을 위하여 차세대 원전 신고리 #3, 4호기에 시설될 WEC의 DCS와 Simulation 서버 그리고 I/O 설비를 구축 중에 있으며, 원자력발전소 현장 기기 모델링 소프트웨어와 I/O 설비간의 인터페이스를 위한 동신 소프트웨어도 개발하고 있다.

• Nuclear Engineering and Technology
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• 제41권5호
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• pp.691-708
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• 2009

In order to analyze thermal hydraulic phenomena during a DVI (Direct Vessel Injection) line break LOCA (Loss-of-Coolant Accident) in the APR1400 (Advanced Power Reactor 1400 MWe), we performed experimental studies with the SNUF (Seoul National University Facility), a reduced-height and reduce-pressure integral test loop with a scaled down APR1400. We performed experiments dealing with eight test cases under varied tests. As a result of the experiment, the primary system pressure, the coolant temperature, and the occurrence time of the downcomer seal clearing were affected significantly by the thermal power in the core and the SI flow rate. The break area played a dominant role in the vent of the steam. For our analytical investigation, we used the MARS code for simulation of the experiments to validate the calculation capability of the code. The results of the analysis showed good and sufficient agreement with the results of the experiment. However, the analysis revealed a weak capability in predicting the bypass flow of the SI water toward the broken DVI line, and it was insufficient to simulate the streamline contraction in the broken side. We, hence, need to improve the MARS code.

• 한국압력기기공학회 논문집
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• 제15권1호
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• pp.18-27
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• 2019

The Advanced Power Reactor 1400 (APR1400) Steam Generator (SG) uses alloy 690 as a tube material and SA-508 Grade 3 Class 1 as a tubesheet material to form tube-to-tubesheet joint through hydraulic expansion process. In this paper, the residual stresses in the SG tube-to-tubesheet contact area was investigated by applying Model-Based System Engineering (MBSE) methodology and the V-model. The use of MBSE transform system description into diagrams which clearly describe the logical interaction between functions hence minimizes the risk of ambiguity. A theoretical and Finite Element Methodology (FEM) was used to assess and compare the residual stresses in the tube-to-tubesheet contact area. Additionally, the axial strength of the tube to tubesheet joint based on the pull-out force against the contact joint force was evaluated and recommended optimum autofrettage pressure to minimize residual stresses in the transition zone given. A single U-tube hole and tubesheet with ligament thickness was taken as a single cylinder and plane strain condition was assumed. An iterative method was used in FEM simulation to find the limit autofrettage pressure at which pull-out force and contact force are of the same magnitude. The joint contact force was estimated to be 20 times more than the pull-out force and the limit autofrettage pressure was estimated to be 141.85MPa.

• 방사성폐기물학회지
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• 제17권4호
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• pp.419-427
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• 2019

APR 1400 액체방사성폐기물관리계통 효율성 증가와 계통의 성능 개선을 위한 방안으로 핵종 이온 광물화 처리기술을 적용하는 것을 고려하였다. 핵종 이온 광물화 처리기술은 현재까지 발전소에 실제적으로 적용되진 않았지만 원자력발전소의 액체방사성폐기물에 존재하는 다양한 핵종 이온을 최소 95% 이상 선택적으로 제거 가능 하다는 것을 실험적으로 증명한 바 있다. 본 논문은 핵종 이온 광물화 처리기술의 제염율을 반영하여 기존 설계에 적용 가능성을 확인하였으며, 기존 설계를 개선할 수 있는 방안을 마련하였다. 핵종 이온 광물화 처리기술의 제염 특성과 기존의 액체방사성폐기물관리계통 설계 및 운전 경험을 고려하여 최적의 적용 위치를 결정하였다. 원자력발전 운영에 따라 발생하는 액체방사성물질이 수집되는 수집탱크에 핵종 이온 광물화 처리기술을 적용하는 것이 기존 설계의 영향이 가장 적을 것이며, 개선 효과도 가장 큰 것으로 해석되었다. 핵종 이온 광물화 처리기술이 현재의 APR 1400 발전소 또는 신규 원전에 적용될 경우 액체방사성폐기물관리계통의 운전 효율성 증가와 계통의 성능 개선이 기대된다.

• 한국안전학회지
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• 제28권3호
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• pp.123-128
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• 2013

The Advanced Power Reactor Plus(APR+), which is a GEN III+ reactor based on the APR1400, is being developed in Korea. In order to enhance the safety of the APR+, a passive auxiliary feedwater system(PAFS) has been adopted in the APR+. The PAFS replaces the conventional active auxiliary feedwater system(AFWS) by introducing a natural driving force mechanism while maintaining the system function of cooling the primary side and removing the decay heat. As the PAFS completely replaces the conventional AFWS, it is required to verify the cooling capacity of PAFS for the core damage frequency(CDF) evaluation. For this reason, this paper discusses the cooling performance of the PAFS during transient accidents. The test case and scenarios were picked from the result of the sensitivity analysis in APR+ Probabilistic Safety Assessment(PSA). The analysis was performed by the best estimate thermal-hydraulic code, RELAP5/.MOD3.3. This study shows that the plant maintains the stable state without the core damages under the given test scenarios. The results of PSA considering this analysis' results shows that the CDF values are decreased. The analysis results can be used for more realistic and accurate performance of a PSA.

• Nuclear Engineering and Technology
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• 제43권3호
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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.

• 시스템엔지니어링학술지
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• 제17권1호
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• pp.21-34
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• 2021

The use of Field-Programmable Gate Arrays (FPGAs) in the development of safety-related Human-Machine Interface (HMI) systems has gained much momentum in nuclear applications. Recently, one of the application areas for the Advanced Power Reactor 1400 (APR1400) is in the development of the advanced Component Interface Module (CIM) of the Engineered Safety Features Actuation System (ESFAS). Using systems engineering approach, we have developed a new FPGA-based advanced CIM software. The first step of our software development process involves the Preliminary Hazard Analysis (PHA) based on the previous CIM design. In this paper, we describe the qualitative approach used in performing the preliminary hazard analysis. The paper presents the methodology for applying a modified Hazard and Operability (HAZOP) procedure for the conduct of PHA which resulted in a qualitative risk-ranking scheme that informed the decisions for the safety criteria in the requirements specification phase. The qualitative approach provided the justification for design changes during the advanced CIM software development process.

• Nuclear Engineering and Technology
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• 제35권2호
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• pp.91-107
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• 2003

Since DBA(Design Basis Accidents) has been studied rather separately from SA(Severe Accidents) in the conventional nuclear reactor safety analysis, the thermal hydraulics during transition between DBA and SA has not been identified so much as each accident itself. Thus, in this study, the thermal hydraulic behavior from DBA to the commencement of SA has been experimentally and analytically investigated for the long-term cooling phase of LB-LOCA(Large-Break Loss-of-Coolant Accident). Experiments were conducted for both cases of the loop seal open and closed in an integral test loop, named as SNUF (Seoul National University Facility), which was scaled down to l/6.4 in length and 1/178 in area of the APR1400 (Advanced Power Reactor 1400MWe). The core mixture level was a main measured value since it took major role in the fuel heat-up rate, the location of fuel melting initiation and the channel blockage by melting material during SA. Experimental results were compared to MAAP4.03 to assess its model of calculating the core mixture level. MAAP4.03 overestimates the core two- phase mixture level because sweep-out and spill-over and the measures to simulate the status of loop seal are not included, which is against the conservatism. Thus, it is recommended that MAAP4.03 should be improved to simulate the thermal hydraulic phenomena, such as sweep-out, spill-over and the status of loop seal.