Static and transient analyses of Advanced Power Reactor 1400 (APR1400) initial core using open-source nodal core simulator KOMODO

Jwaher Alnaqbi; Donny Hartanto; Reem Alnuaimi; Muhammad Imron; Victor Gillette

Nuclear Engineering and Technology (Feb 2022)

Static and transient analyses of Advanced Power Reactor 1400 (APR1400) initial core using open-source nodal core simulator KOMODO

Jwaher Alnaqbi,
Donny Hartanto,
Reem Alnuaimi,
Muhammad Imron,
Victor Gillette

Affiliations

Jwaher Alnaqbi: New York University Abu Dhabi, P.O. BOX 129188, Abu Dhabi, United Arab Emirates
Donny Hartanto: Department of Mechanical and Nuclear Engineering, University of Sharjah, P.O. BOX 27272, Sharjah, United Arab Emirates; Nuclear Energy System Simulation and Safety Research Group, Research Institute of Sciences and Engineering, University of Sharjah, P.O. BOX 27272, Sharjah, United Arab Emirates; Corresponding author. Department of Mechanical and Nuclear Engineering, University of Sharjah, P.O. BOX 27272, Sharjah, United Arab Emirates.
Reem Alnuaimi: Department of Mechanical and Nuclear Engineering, University of Sharjah, P.O. BOX 27272, Sharjah, United Arab Emirates
Muhammad Imron: Department of Nuclear Engineering, Ulsan National Institute of Science and Technology, 50 UNIST-gil, Eonyang-eup, Ulju-gun, Ulsan, 44919, Republic of Korea
Victor Gillette: Department of Mechanical and Nuclear Engineering, University of Sharjah, P.O. BOX 27272, Sharjah, United Arab Emirates; Nuclear Energy System Simulation and Safety Research Group, Research Institute of Sciences and Engineering, University of Sharjah, P.O. BOX 27272, Sharjah, United Arab Emirates

Journal volume & issue: Vol. 54, no. 2
pp. 764 – 769

Abstract

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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.

Published in Nuclear Engineering and Technology

ISSN: 1738-5733 (Print)
Publisher: Elsevier
Country of publisher: Korea, Republic of
LCC subjects: Technology: Electrical engineering. Electronics. Nuclear engineering: Nuclear engineering. Atomic power
Website: http://www.journals.elsevier.com/nuclear-engineering-and-technology

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