Steady-State Thermal-Hydraulic Analysis of TRIGA Research Reactor  

作　　者：Mohammad Mizanur Rahman Mohammad Abdur R. Akond Mohammad Khairul Basher Md. Quamrul Huda 

机构地区：[1]Nuclear Energy Division, Energy Institute, Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh [2]Renewable Energy Division, Energy Institute, Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh

出　　处：《World Journal of Nuclear Science and Technology》2014年第2期81-87,共7页核科学与技术国际期刊（英文）

摘　　要：The COOLOD-N2 and PARET computer codes were used for a steady-state thermal hydraulic and safety analysis of the 3 MW TRIGA Mark-II research reactor located at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh. The objective of the present study is to ensure that all important safety related thermal hydraulic parameters uphold margins far below the safety limits by steady-state calculations at full power. We, therefore, have calculated the hot channel fuel centreline temperature, fuel surface temperature, cladding surface temperature, the departure from nucleate boiling (DNB) heat flux and DNB ratio, axial fuel centreline temperature and compared. The comparison indicates that the calculated values are in satisfactory agreement between the codes. The data obtained in this investigation are largely far to compromise safety of the reactor. The results can also be used to upgrade the current core configuration of the TRIGA reactor.The COOLOD-N2 and PARET computer codes were used for a steady-state thermal hydraulic and safety analysis of the 3 MW TRIGA Mark-II research reactor located at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh. The objective of the present study is to ensure that all important safety related thermal hydraulic parameters uphold margins far below the safety limits by steady-state calculations at full power. We, therefore, have calculated the hot channel fuel centreline temperature, fuel surface temperature, cladding surface temperature, the departure from nucleate boiling (DNB) heat flux and DNB ratio, axial fuel centreline temperature and compared. The comparison indicates that the calculated values are in satisfactory agreement between the codes. The data obtained in this investigation are largely far to compromise safety of the reactor. The results can also be used to upgrade the current core configuration of the TRIGA reactor.

关 键 词：COOLOD-N2 PARET TRIGA Mark-II DNB Safety 

分 类 号：R73[医药卫生—肿瘤]