• Journal of the Korean Society of Systems Engineering
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• v.13 no.1
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• pp.15-23
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• 2017

This paper describes verification of APR1400 reactor vessel by applying the system engineering approach, in which the design re-engineering method is used to check the design parameters of APR1400 RV (reactor vessel). The RV is classified as safety class 1 and therefore must adhere strictly to the rules of ASME BPVC section III, subsection NB and seismic category I. This study explores designing the RV by following the ASME guidelines and making a comparative study with the current design. To meet this objective we apply system engineering methodologies to structure the process and allow for verification and validation of the major RV design parameters such as thickness of RV. The structural thicknesses of various part of RV are determined as well as reinforcements on the RV major nozzles. A 3D virtual reality model was created based on the design parameters using CATIA V5 and animation using Dassault Composer V2016. A comparison of re-engineered ARP1400 RV and standard APR1400 RV was done to show which design parameters were taken more conservative approach.

• 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.

• Proceedings of the KIEE Conference
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• 2001.07a
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• pp.87-89
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• 2001

On-site power system design of APR1400 is different from that of existing and operating plants and APR1400 has no operating experience. So we have to confirm its adequacy of design exclusively by analyses. So an method of analysis is the only way to evaluate safety of design of the power system of APR1400. Therefore the purpose of this paper is a construction of verification program and a verification of utilities' analysis results by using this program in order to confirm the adequacy of APR1400 on-site & off-site power-system.

• 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.

• 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.

• Korean Journal of Construction Engineering and Management
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• v.20 no.5
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• pp.52-60
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• 2019

The APR1400 is a 1400MWe nuclear power plant developed through national technology development project over a period about 10years. Approximately 65,000 design drawings are produced for APR1400 construction. In order to maintain consistency among numerous drawings, the highest level of design bases drawings (DBDs) are created according to design bases and this is used in the subsequent design. However, DBDs are produced and managed on a document basis and they are managed various field, it was difficult to accurately reflect the design bases information in the subsequent design. Therefore, this study recognizes limitations of the document based DBDs and develops a system that can accurately reflect the design bases information to subsequent design by adopting BIM based design bases integrated information system. Especially, by introducing DBIL(Design Bases Information Layer) concept, DBIL was created and analyzed based on five design bases(Physical protection, Fire protection, Internal missile protection, Internal flood protection, Radiation protection) applied to APR1400. In the final result DBIL set and Datasheet are integrated of room, design bases information, building data(wall, slab, door, window, penetrations). So it can be used for subsequent design automation and design verification. Furthermore, it is expected that APR1400 DBILs data can be used extensively in constructability and design economics analysis through comparison with next generation nuclear power plant.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.56.2-56.2
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.171-171
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• 2002

• 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).

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.869-877
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• 2020

A 24-month Advanced Power Reactor 1400 (APR1400) core with a very-low-boron (VLB) concentration has been investigated for an inherently safe and high-performance PWR in this work. To develop a high-performance APR1400 which is able to do the passive frequency control operation, VLB feature is essential. In this paper, the centrally-shielded burnable absorber (CSBA) is utilized for an efficient VLB operation in the 24-month cycle APR1400 core. This innovative design of the VLB APR1400 core includes the optimization of burnable absorber and loading pattern as well as axial cutback for a 24-month cycle operation. In addition to CSBA, an Er-doped guide thimble is also introduced for partial management of the excess reactivity and local peaking factor. To improve the neutron economy of the core, two alternative radial reflectors are adopted in this study, which are SS-304 and ZrO₂. The core reactivity and power distributions for a 2-batch equilibrium cycle are analyzed and compared for each reflector design. Numerical results show that a VLB core can be successfully designed with 24-month cycle and the cycle length is improved significantly with the alternative reflectors. The neutronic analyses are performed using the Monte Carlo Serpent code and 3-D diffusion code COREDAX-2 with the ENDF/B-VII.1.