Development of advanced power reactor nuclear power plants containment pressure and temperature analysis methodology using CAP computer code

IF 1.5 4区 工程技术 Q3 ENGINEERING, MECHANICAL
Yong-Ju Cho, Sun-Chang Moon, Dae-Hyung Lee, Sun-Hong Yoon
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Abstract

The paper provides a detailed overview of the development of a containment pressure and temperature (P/T) analysis methodology for advanced power reactor 1400 (APR1400) nuclear power plants (NPPs). The study addresses the restrictions on exporting independent nuclear power plants by utilizing the containment analysis package (CAP) computer code developed in Korea. One of the key aspects highlighted in the paper is the comparison of results obtained from the CAP code with those from the CONTEMPT-LT/028 code which is used P/T analysis. The analysis focuses on two types of accidents: loss of coolant accidents (LOCA) and main steam line break (MSLB) accidents, which are considered as design basis accidents. By comparing the outcomes of both codes, the paper evaluates the performance and effectiveness of the CAP code in predicting the P/T behavior within the containment during these accidents. The paper also discusses the criteria and technical standards for the containment P/T analysis. It emphasizes the importance of ensuring that the peak P/T remain within the safety related systems, equipment, and structures of NPP. The design pressure is identified as a critical factor in achieving these objectives. In conclusion, the study presents the successful development of a containment P/T analysis methodology using the CAP computer code for APR1400. The methodology considers the specific characteristics of Korean NPPs and new code, CAP. The paper emphasizes the applicability and effectiveness of the CAP code in this context. However, further research and validation efforts are recommended to enhance the accuracy and reliability of the methodology for various design basis accidents. The developed methodology is expected to contribute to the safe and efficient operation of APR1400 NPPs and support Korea’s ambitions in exporting NPPs’ technology to other countries.

利用 CAP 计算机代码开发先进动力反应堆核电站安全壳压力和温度分析方法
本文详细概述了先进动力反应堆 1400(APR1400)核电站(NPP)安全壳压力和温度(P/T)分析方法的开发情况。该研究利用韩国开发的安全壳分析软件包 (CAP) 计算机代码,解决了独立核电站出口的限制问题。本文强调的一个关键方面是将 CAP 代码与用于 P/T 分析的 CONTEMPT-LT/028 代码的结果进行比较。分析的重点是两类事故:冷却剂损失事故 (LOCA) 和主蒸汽管线断裂事故 (MSLB),这两类事故被视为设计基础事故。通过比较两种规范的结果,本文评估了 CAP 规范在预测这些事故期间安全壳内 P/T 行为方面的性能和有效性。论文还讨论了安全壳 P/T 分析的标准和技术规范。它强调了确保峰值 P/T 保持在核电厂安全相关系统、设备和结构内的重要性。设计压力是实现这些目标的关键因素。总之,本研究介绍了使用 APR1400 的 CAP 计算机代码成功开发的安全壳 P/T 分析方法。该方法考虑了韩国核电站和新代码 CAP 的具体特点。本文强调了 CAP 代码在此背景下的适用性和有效性。不过,建议进一步开展研究和验证工作,以提高该方法在各种设计基础事故中的准确性和可靠性。预计所开发的方法将有助于 APR1400 核电站的安全高效运行,并支持韩国向其他国家出口核电站技术的雄心。
本文章由计算机程序翻译,如有差异,请以英文原文为准。
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来源期刊
Journal of Mechanical Science and Technology
Journal of Mechanical Science and Technology 工程技术-工程:机械
CiteScore
2.90
自引率
6.20%
发文量
517
审稿时长
7.7 months
期刊介绍: The aim of the Journal of Mechanical Science and Technology is to provide an international forum for the publication and dissemination of original work that contributes to the understanding of the main and related disciplines of mechanical engineering, either empirical or theoretical. The Journal covers the whole spectrum of mechanical engineering, which includes, but is not limited to, Materials and Design Engineering, Production Engineering and Fusion Technology, Dynamics, Vibration and Control, Thermal Engineering and Fluids Engineering. Manuscripts may fall into several categories including full articles, solicited reviews or commentary, and unsolicited reviews or commentary related to the core of mechanical engineering.
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