• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.764-769
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• 2022

The United Arab Emirates is currently building and operating four units of the APR-1400 developed by a South Korean vendor, Korea Electric Power Corporation (KEPCO). This paper attempts to perform APR-1400 reactor core analysis by using the well-known two-step method. The two-step method was applied to the APR-1400 first cycle using the open-source nodal diffusion code, KOMODO. In this study, the group constants were generated using CASMO-4 fuel transport lattice code. The simulation was performed in Hot Zero Power (HZP) at steady-state and transient conditions. Some typical parameters necessary for the Nuclear Design Report (NDR) were evaluated in this paper, such as effective neutron multiplication factor, control rod worth, and critical boron concentration for steady-state analysis. Other parameters such as reactivity insertion, power, and fuel temperature changes during the Reactivity Insertion Accident (RIA) simulation were evaluated as well. The results from KOMODO were verified using PARCS and SIMULATE-3 nodal core simulators. It was found that KOMODO gives an excellent agreement.

• 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.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2498-2515
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• 2024

In this study, we propose a fully parallelized adaptive particle refinement (APR) algorithm for smoothed particle hydrodynamics (SPH) to construct a stable and efficient multi-resolution computing system for nuclear safety analysis. The APR technique, widely employed by SPH research groups to adjust local particle resolutions, currently operates on a serialized algorithm. However, this serialized approach diminishes the computational efficiency of the system, negating the advantages of acceleration achieved through high-performance computing devices. To address this drawback, we propose a fully parallelized APR algorithm designed to enhance both efficiency and computational accuracy, facilitated by a new adaptive smoothing length model. For model validation, we simulated both hydrostatic and hydrodynamic benchmark cases in 2D and 3D environments. The results demonstrate improved computational efficiency compared to the conventional SPH method and APR with a serialized algorithm, and the model's accuracy was confirmed, revealing favorable outcomes near the resolution interface. Through the analysis of jet breakup, we verified the performance and accuracy of the model, emphasizing its applicability in practical nuclear safety analysis.

• Proceedings of the KIEE Conference
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• 2008.10b
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• pp.411-412
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• 2008

본 논문에서는 차세대원자로인 APR1400(Advanced Power Reactor 1400)의 출력제어방법으로 모델예측제어 알고리즘을 적용하고, 일일부하추종 운전을 하였을 때 최적의 제어기 구현을 위해 제어 로직의 주요 변수인 예측구간, 제어구간, 모델 차수의 변화에 따른 제어 성능을 평가하였다. 성능 평가는 원자로 출력제어 성능 검증시 사용하는 방법으로 제어대상인 차세대 원자로(APR1400)를 3차원 노심해석 전산코드인 MASTER(Multipurpose Analyzer for Static and Transient Effects of Reactor)로 시뮬레이션하여 제어 성능을 평가하였다.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.21 no.8
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• pp.708-713
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• 2011

U.S. nuclear regulatory commission(NRC) regulatory guide(RG) 1.20 requires a comprehensive vibration assessment program(CVAP) for use in verifying the structural integrity of reactor vessel internals(RVI) for flow-induced vibrations prior to commercial operation. The CVAP program consist of vibration and fatigue analysis, a vibration measurement program, an inspection program, and a correlation of their results. One of the main purposes of the analysis program is to select measurement locations, however measurement locations can not be determined by only analysis results, therefore we developed selection criteria of measurement locations for advanced power reactor 1400(APR1400) RVI CVAP, It will be used to select measurement locations and instrument types for APR1400 RVI CVAP.

• 한국IT서비스학회:학술대회논문집
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• 2009.05a
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• pp.255-259
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• 2009

해외시장 진출이 가능한 국내 고유노형 개발을 목표로 지난 '07년 8월부터 APR+(Advanced Power Reactor Plus)기술개발 사업이 추진 중에 있다. 본 사업을 통하여 생산 되는 원전 설계 결과물들을 기존의 파일 기반 관리 시스템에서 진일보한 데이터 기반 관리시스템을 개발하여 관리하고자 한다. 본 시스템은 원자력 발전소 전 수명주기 동안 데이터 간의 유기적 연계 사용을 목표로 하고 국제표준을 사용하여 개방형 시스템으로 구축한다. 본 연구는 APR+ 설계정보관리시스템 구축을 위한 기반연구로써 국제표준 기술 및 원자력 발전 분야의 정보관리 사례 등을 분석하여 시스템 개발 방향을 전망하고자 한다.

• Journal of the Korean Society of Safety
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• v.16 no.3
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• pp.125-134
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• 2001

The Advanced Pressurized rater Reactor 1400(APR1400) system is advanced of the successful Korean Nuclear Power Plants(KSNP) design which meets functional needs for safety enhancement reliability improvement, and control in the human-computer monitoring system. Therefore this paper describes the scoring model in order to justify the reliability and safety in APR 1400 under uncertainty. The structure of this paper consists of the human engineering, risk safety, quality function, safety organization management factors of the qualitative factors in chapter 2, and the expectation results of the normalized scoring model in chapter 3. Finally, the proposed reliability model have provided the technical flexibility not only for functional control fields but also for accidents protection systems in APR 1400 under uncertainty.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.2047-2052
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• 2004

Inconel-600 alloy has been used as steam generator tube material for current pressurized water reactors (PWRs). The long-term operation of steam generators showed that the use of this material induced localized corrosion damages and increased tube wear of steam generator. To protect these problems, steam generator tube material is being changed to Inconel-690 alloy. Based on the current trend, we have chosen Inconel 690 as the Advanced Power Reactor 1400 (APR1400) steam generator(SG) tube material and performed the design optimization of preventive measure against tube fretting wear for the APR1400 steam generator. In this paper, we examined the technical consideration in this modification : the selection of material, wear characteristics, effect of the Egg-crate Flow Distribution Plate installation, and effect analysis of vertical strip installation.

• Journal of the KSME
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• v.52 no.3
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• pp.28-31
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• 2012

이 글에서는 한국수력원자력(주) 주관으로 개발 중인 차세대 원자력발전소(APR+)에 대해 기본적인 개념과 개발진행 현황에 대해 소개하고자 한다.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.128.1-128
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• 2001