应用 EUCLID 集成代码的 HYDRA-IBRAE/LM 热液压模块分析钠冷反应堆发电厂的蒸汽发生器

IF 0.9 Q4 ENERGY & FUELS
I. A. Klimonov, N. A. Mosunova, V. F. Strizhov, E. V. Usov, V. I. Chukhno
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引用次数: 0

摘要

摘要--应用基于现代物理和数学模型的计算工具来证实各种传热设备组件所采用的设计方案,有助于节省设计机构的时间、人力和财力。现有和正在设计的反应堆种类繁多,在设计和冷却剂类型上都各不相同,这就需要有一种适用于广泛应用的多功能热工水力计算机代码。作为 Proryv(突破)项目的一部分而开发的 EUCLID 集成代码的新一代 HYDRA-IBRAE/LM 热液压模块可以满足这些要求。作为集成代码的一部分,该热工水力模块的运行为模拟范围更广的反应堆厂房运行模式提供了可能,从而也为模拟单个换热设备组件的运行模式提供了可能。所开发的热工水力模块已通过核与辐射安全科学与工程中心(SEC NRS)的认证,可以分析核电站各种设备中钠、铅、铅铋、气体和水冷却剂的热工水力。反应堆厂房蒸汽发生器(SG)属于建模最复杂的设备部件,因为它们可能包含两种冷却剂。文章介绍的研究结果表明,代码能够以正确的方式分析钠冷却反应堆厂房蒸汽发生器中的过程,因为这些厂房在俄罗斯和世界各地都存在并在积极运行。根据文章中提供的数据可以得出结论,IBRAE RAS 开发的热工水力模块是对反应堆厂房复杂传热过程进行数值分析的有效工具。通过使用模块中的扩展闭合相关系统,可以对应用于单个传热设备组件的热工设计方案进行验证。
本文章由计算机程序翻译,如有差异,请以英文原文为准。

Application of the EUCLID Integrated Code’s HYDRA-IBRAE/LM Thermal Hydraulic Module for Analyzing the Steam Generators of Sodium Cooled Reactor Plants

Application of the EUCLID Integrated Code’s HYDRA-IBRAE/LM Thermal Hydraulic Module for Analyzing the Steam Generators of Sodium Cooled Reactor Plants

Abstract—

Application of computation tools resting on contemporary physical and mathematical models for substantiating the design solutions adopted for various heat-transfer equipment components helps save time, manpower, and financial resources of design institutions. The variety of both existing reactors and those being designed, which differ from one another both in design and type of coolants calls for the availability of a versatile thermal hydraulic computer code suited for a wide range of applications. The new-generation HYDRA-IBRAE/LM thermal hydraulic module of the EUCLID integrated code, which has been developed as part of the Proryv (Breakthrough) Project, meets these requirements. The operation of this thermal hydraulic module as part of the integrated code opens the possibility to simulate an essentially wider range of reactor plant operation modes and, as a consequence, those of individual heat-transfer equipment components. The developed thermal hydraulic module, which has been certified at the Scientific and Engineering Center for Nuclear and Radiation Safety (SEC NRS), offers the possibility to analyze the thermal hydraulics of sodium, lead, lead–bismuth, gas, and water coolants in various NPP equipment items. Reactor plant steam generators (SGs) belong to the category of equipment components most complex for modeling since they may contain two types of coolants. The article presents study results demonstrating the code’s abilities to analyze in a correct way the processes in the steam generators of only sodium cooled reactor plants, because these plants exist and are actively operated in Russia and around the world. The data presented in the article allow a conclusion to be drawn that the thermal hydraulic module developed at IBRAE RAS is an efficient tool for numerically analyzing complex heat-transfer processes in reactor plants. By using an extended system of closing correlations implemented in the module, it is possible to perform substantiation of design thermal engineering solutions as applied to individual heat-transfer equipment components.

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来源期刊
CiteScore
1.30
自引率
20.00%
发文量
94
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