NUMERICAL STUDY ON THERMAL BEHAVIOUR AND MODERATOR BOILING IN TWO NUCLEAR CHANNELS OF IPHWR 220 MWE REACTOR AT DIFFERENT HEAT INPUTS DURING LOSS OF COOLANT ACCIDENT (LOCA)

IF 2.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy Pub Date : 2025-09-01 DOI:10.1016/j.anucene.2025.111847

Raushan Kumar, Chandranshu Kumar Singh, Anurag Singh, Mushtaque Momin, Mukesh Sharma

{"title":"NUMERICAL STUDY ON THERMAL BEHAVIOUR AND MODERATOR BOILING IN TWO NUCLEAR CHANNELS OF IPHWR 220 MWE REACTOR AT DIFFERENT HEAT INPUTS DURING LOSS OF COOLANT ACCIDENT (LOCA)","authors":"Raushan Kumar, Chandranshu Kumar Singh, Anurag Singh, Mushtaque Momin, Mukesh Sharma","doi":"10.1016/j.anucene.2025.111847","DOIUrl":null,"url":null,"abstract":"<div><div>Indian PHWRs, derived from CANDU designs, are vulnerable to postulated accident scenarios like LOCA with simultaneous ECCS failure. To investigate such conditions, this study analyses the temperature distribution in the pressure tube (PT) and calandria tube (CT) under voided channel conditions using steady-state thermal analysis and ANSYS Fluent at various heat fluxes (900, 1500, 2500, 3500, and 4500 W/m<sup>2</sup>). The PT showed negligible circumferential temperature variation, while the CT exhibited significant heating near neighbouring channels. As heat flux increased, the moderator surrounding the CT progressively heated and eventually began to boil beyond 1500 W/m<sup>2</sup>. Boiling was more intense near the lower region of channel 1 and the upper region of channel 2. The trend in boiling behaviour remained consistent across different heat fluxes, although the onset occurred earlier at higher flux levels. These findings highlight critical thermal behaviour under accident-like conditions and can support safety evaluations of reactor designs.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111847"},"PeriodicalIF":2.3000,"publicationDate":"2025-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":"0","resultStr":null,"platform":"Semanticscholar","paperid":null,"PeriodicalName":"Annals of Nuclear Energy","FirstCategoryId":"5","ListUrlMain":"https://www.sciencedirect.com/science/article/pii/S0306454925006644","RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":null,"EPubDate":"","PubModel":"","JCR":"Q1","JCRName":"NUCLEAR SCIENCE & TECHNOLOGY","Score":null,"Total":0}

引用次数: 0

Abstract

Indian PHWRs, derived from CANDU designs, are vulnerable to postulated accident scenarios like LOCA with simultaneous ECCS failure. To investigate such conditions, this study analyses the temperature distribution in the pressure tube (PT) and calandria tube (CT) under voided channel conditions using steady-state thermal analysis and ANSYS Fluent at various heat fluxes (900, 1500, 2500, 3500, and 4500 W/m²). The PT showed negligible circumferential temperature variation, while the CT exhibited significant heating near neighbouring channels. As heat flux increased, the moderator surrounding the CT progressively heated and eventually began to boil beyond 1500 W/m². Boiling was more intense near the lower region of channel 1 and the upper region of channel 2. The trend in boiling behaviour remained consistent across different heat fluxes, although the onset occurred earlier at higher flux levels. These findings highlight critical thermal behaviour under accident-like conditions and can support safety evaluations of reactor designs.

查看原文本刊更多论文

失冷事故（局部）中不同热输入下iphwr 220mwe反应堆两个核通道的热行为和慢化剂沸腾数值研究

来自CANDU设计的印度phwr很容易受到LOCA等假想事故的影响，同时ECCS发生故障。为了研究这种情况，本研究利用稳态热分析和ANSYS Fluent软件，在不同的热流密度（900、1500、2500、3500和4500 W/m2）下，分析了空通道条件下压力管（PT）和万向管（CT）的温度分布。PT显示可以忽略的周向温度变化，而CT显示邻近通道附近明显的加热。随着热流密度的增加，CT周围的慢化剂逐渐升温，最终开始沸腾，超过1500 W/m2。在通道1的下部和通道2的上部附近沸腾更强烈。沸腾行为的趋势在不同的热通量上保持一致，尽管在高通量水平上发生得较早。这些发现强调了类似事故条件下的临界热行为，可以支持反应堆设计的安全性评估。

本文章由计算机程序翻译，如有差异，请以英文原文为准。

求助全文

约1分钟内获得全文求助全文

来源期刊

Annals of Nuclear Energy 工程技术-核科学技术

CiteScore

4.30

自引率

21.10%

发文量

632

审稿时长

7.3 months

期刊介绍： Annals of Nuclear Energy provides an international medium for the communication of original research, ideas and developments in all areas of the field of nuclear energy science and technology. Its scope embraces nuclear fuel reserves, fuel cycles and cost, materials, processing, system and component technology (fission only), design and optimization, direct conversion of nuclear energy sources, environmental control, reactor physics, heat transfer and fluid dynamics, structural analysis, fuel management, future developments, nuclear fuel and safety, nuclear aerosol, neutron physics, computer technology (both software and hardware), risk assessment, radioactive waste disposal and reactor thermal hydraulics. Papers submitted to Annals need to demonstrate a clear link to nuclear power generation/nuclear engineering. Papers which deal with pure nuclear physics, pure health physics, imaging, or attenuation and shielding properties of concretes and various geological materials are not within the scope of the journal. Also, papers that deal with policy or economics are not within the scope of the journal.