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In this paper, a novel theoretical model has been developed to obtain focal points of temperatures in helium cooled ceramic breeder (HCCB) blanket system of China Fusion Engineering Test Reactor (CFETR). The main objective of the paper is to develop a theoretical-computational model for safety of blanket system, structure stress analysis and feasibility design. The theoretical-computational model was developed and verified by computational techniques using C++/MATLAB program and later with thermo-hydraulic system codes, i.e. RELAP5 and THEATRE. Mathematical model was developed for the thermal conductivity in HCCB blanket system and heat conduction model by using numerical technique. This model was then transferred to C++/MATLAB program and developed a program for calculating the thermal contact temperature of HCCB region. The calculated temperature values were then compared with RELAP5 and THEATRE codes for the accuracy of theoretical model. The preliminary thermal-hydraulic simulation results indicate that the maximum temperature on the first wall (FW) satisfies the design allowable limit and that the cooling channels were designed accurately. The results obtained from these analyses are very close to the experimentally defined values depicting the accuracy of theoretical and simulation results. The calculated results can be applicable to other tokamak devices like CFETR, ITER and JET as well. [ABSTRACT FROM AUTHOR]

Thermal quenching in Tokamak reactor is the most obvious phenomenon happens during plasma disruption conditions. The current research is focused on the thermal behavior of different parameters of China fusion engineering test reactor (CFETR) including reactor power, pressure and mass flow rate conditions. The analysis was performed by two thermal hydraulic codes, i.e. THEATRe and Relap5. During the first phase of research, thermal quenching behavior and trends that can be possible during the reactor operation was performed. In the next phase, nodalization diagram of THEATRe and Relap5 codes were developed. The listed parameters were calculated and analyzed for the safety aspects of the reactor. The main objective of the research was to analyze the blanket system of CFETR (Tokamak) for safety concerns during disruption condition. The research will be extended to other components for safe operation of reactor as well. [ABSTRACT FROM AUTHOR]