Subchannel Analysis of Critical and Subcritical Core of a High Flux Research Reactor
Samiran Sengupta1,a, Tej Singh2, P. K. Vijayan3, K. Sasidhran4 and V. K. Raina1,b
1Research Reactor Design & Projects Division, Bhabha Atomic Research Centre, Mumbai 400 085, India.
asamiran_sengupta@yahoo.co.in
bksasi@barc.gov.in
2Research Reactor Services Division, Bhabha Atomic Research Centre, Mumbai 400 085, India.
t_singh@barc.gov.in
3Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai 400 085, India.
vijayanp@barc.gov.in
4Reactor Group, Bhabha Atomic Research Centre, Mumbai 400 085, India.
vkrain@barc.gov.in
ABSTRACT
This paper presents subchannel analysis of both critical and subcritical core of a high flux research reactor using computer code COBRA-IV. Various parameters related to fuel and coolant channels such as flow distribution in the coolant channels, coolant velocities, fuel rod central temperature, clad surface temperature, onset of nucleate boiling temperature, departure from nucleate boiling heat flux, onset of nucleate boiling temperature margin and departure from nucleate boiling ratio for each fuel element are computed using this code. Methodology & results of the two levels of calculation using i) assembly level subchannel analysis and ii) single closed hottest subchannel analysis are presented. Various correlations applicable for low pressure and high mass flux conditions of research reactor are used to calculate departure from nucleate boiling heat flux. The details of the modeling and results are described in this paper.
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