Analysis of double-ended guillotine break accident of surge line in ACP100 based on coupling method: Development and validation of the coupling method
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The ACP100 is a small modular reactor (SMR) designed and built in China, featuring an integrated primary loop and passive safety systems. In SMRs, the interaction between the reactor coolant system, containment vessel, and other systems is highly interdependent. Therefore, it is necessary to use high-precision analysis codes, which can be achieved by coupling multiple codes. This paper is the first part of a study which investigates the LOCA in the ACP100 using a coupled platform. In this paper, a coupling platform was developed for transient analysis of SMR accidents, based on the system code NUSOL-SYS and the Integrated Severe Accident Analysis (ISAA) code. An inter-process communication module was developed adopting shared memory and event objects to exchange data between the two codes. A coupling interface defining the data to be exchanged was proposed. Both NUSOLSYS and ISAA were modified to perform synchronous time step control. The coupling platform were validated through a hypothetical scenario and Edwards’ pipe blowdown experiment, demonstrating exact consistency and high accuracy. This coupling platform offers a new method for SMR accident analysis, providing a foundation for future work
Abstract
The ACP100 is a small modular reactor (SMR) designed and built in China, featuring an integrated primary loop and passive safety systems. In SMRs, the interaction between the reactor coolant system, containment vessel, and other systems is highly interdependent. Therefore, it is necessary to use high-precision analysis codes, which can be achieved by coupling multiple codes. This paper is the first part of a study which investigates the LOCA in the ACP100 using a coupled platform. In this paper, a coupling platform was developed for transient analysis of SMR accidents, based on the system code NUSOL-SYS and the Integrated Severe Accident Analysis (ISAA) code. An inter-process communication module was developed adopting shared memory and event objects to exchange data between the two codes. A coupling interface defining the data to be exchanged was proposed. Both NUSOLSYS and ISAA were modified to perform synchronous time step control. The coupling platform were validated through a hypothetical scenario and Edwards’ pipe blowdown experiment, demonstrating exact consistency and high accuracy. This coupling platform offers a new method for SMR accident analysis, providing a foundation for future work
참고문헌 (21)
- Vaya Soler Antonio, "Small Modular Reactors : Challenges and Opportunities", NEA No. 7560, ~
- H. Boado Magan, "CAREM project status", Science and Technology of Nuclear Installations, 2011, pp.1~6 (2011)[DOI]
- D. Song, "Key technology of ACP100: reactor core and safety design", Nucl. Power Eng, 42(4), pp.1~5 (2021)
- H.-K. Kim, "Thermal-hydraulic analysis of SMART steam generator tube rupture using TASS", SMR-S code, 2013, pp.55, 331~340
- Jos´e N. Reyes, "NuScale plant safety in response to extreme events", Nucl. Technol, 178(2), pp.153~163 (2012)[DOI]
- J. Deng, "Analysis of post-LOCA long-term core safety characteristics for the Small Modular Reactor ACP100", Ann. Nucl. Energy, 142, pp.107349~ (2020)[DOI]
- Q. Xiong, "Design for ACP100 long term cooling flow resistance with random forests and inverse quantification", Ann. Nucl. Energy, 180, pp.109477~ (2023)[DOI]
- Katarzyna Skolik, "Chris Allison, LOCA-Type Scenario Simulation for NuScale-SMR with RELAP", SCDAPSIM, pp.2019, ~
- Mirco Di Giuli, "Exploratory Studies of Small Modular Reactors Using the ASTEC Code", ~ (2015)
- Andrea Achilli, "SPES3 facility RELAP5 sensitivity analyses on the containment system for design review", Science and Technology of Nuclear Installations, 2012, pp.1~19 (2012)[DOI]
- J. Shan, "Development of multi-scale high-fidelity system analysis code NUSOL-SYS", Presented at the CDNET-2018 (2018)
- B. Zhang, "Performance evaluation of Accident Tolerant Fuel under station blackout accident in PWR nuclear power plant by improved ISAA code", Nucl. Eng. Technol, 54(7), pp.2475~2490 (2022)[DOI]
- F. Chao, "Study on the algorithm for solving two-fluid seven-equation two-pressure model", Ann. Nucl. Energy, 111, pp.379~390 (2018)[DOI]
- M.J. Thurgood, "COBRA", TRAC -A Thermal-Hydraulics Code for Transient Analysis of Nuclear Reactor Vessels and Primary Coolant Systems, CR-3046, pp.1983[DOI]
- P.O. Box, Using the Coupled MELCOR-RELAP5 Codes for Simulation of the Edward’s Pipe, "Sal B. Rodríguez", Sandia National Laboratories, ~, pp.Nr. SAND2002-2828C (2002)
- J.A. Turner, "The virtual environment for reactor applications (VERA): design and architecture", J. Comput. Phys, 326, pp.544~568 (2016)[DOI]
- J.R. Lee, "Transient full core analysis of PWR with multi-scale and multi-physics approach", Nucl. Eng. Technol, 56(3), pp.980~992 (2024)[DOI]
- A. Kursad Muftuoglu, "Improved Methodology for Integral Analysis of Advanced Reactors Employing Passive Safety", Ph. D, pp.2004
- R. K. Cole, "P.O. Box, Coupling of MELCOR to other codes under an executive program using PVM message exchange", Presented at the RELAP5 Users Seminar (2002)
- W.L. Weaver, "A PVM executive program for use with RELAP5-3D", 10th International Conference on Nuclear Engineering, 3, pp.1057~1061 (2002), ASMEDC
- A.R. Edwards, "Studies of phenomena connected with the depressurization of water reactors", J. Br. Nucl. Energy Soc, 9, pp.125~135 (1970)
