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Hot channel analysis of a 333 mwth civil nuclear marine core using the cobra-en code


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Alam, SB 
Lindley, BA 
Parks, GT 


In this study, hot channel analysis of a 333 MWth civil nuclear marine Pressurized Water Reactor (PWR) core operating in steady-state conditions has been performed to determine whether it satisfies thermal-hydraulic (TH) safety limits. For this purpose, we have used the code COBRA-EN with standard and the tightest achievable fuel dimensions (lower pitch-to-diameter ratio (P=D)) and typical PWR TH conditions. The reactor power distribution was computed using the WIMS and PANTHER reactor physics codes. The analysis shows that even in the hot channel at 118% overpower, the minimum departure from nucleate boiling ratio (MDNBR) and the fuel temperatures remain well within TH margins for both the standard and low P=D geometry cores. It is also necessary to prevent boiling in the coolant. In the COBRA-EN model, the coolant does not begin to boil unless the core-averaged linear power rating exceeds 27 kW/m for standard geometry, which is 155% higher than the design value. At steady state, due to the increased pressure drop, a low P=D lattice leads to a reduced coolant ow rate. In turn, this leads to a higher temperature rise across the core, which affects temperature limits and the MDNBR. Nevertheless, we find that it is possible to increase the power density by more than 45% while remaining within TH limits.



COBRA-EN, hot channel analysis, DNBR, civil nuclear marine core

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International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015

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American Nuclear Society

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