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.

阿拉伯联合酋长国目前正在建造和运营由韩国供应商韩国电力公司 (KEPCO) 开发的四台 APR-1400。本文尝试使用众所周知的两步法进行 APR-1400 反应堆堆芯分析。使用开源节点扩散代码 KOMODO 将两步法应用于 APR-1400 的第一个周期。在这项研究中，群常数是使用 CASMO-4 燃料输运晶格代码生成的。仿真是在稳态和瞬态条件下的热零功率 (HZP) 中进行的。本文评估了核设计报告 (NDR) 所需的一些典型参数，例如有效中子倍增因子、控制棒价值和用于稳态分析的临界硼浓度。其他参数，如反应性插入、功率、还评估了反应性插入事故 (RIA) 模拟期间的燃料温度变化。使用 PARCS 和 SIMULATE-3 节点核心模拟器验证了 KOMODO 的结果。发现 KOMODO 给出了很好的一致性。