Browsing by Author "Ariman, N."

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    Modelling of Hottest Channels of VVER-1000 and VVER-1200 Nuclear Reactors by Using COBRA-TF and ZEBRA Thermal Hydraulic Codes
    (Ankara Üniversitesi Nükleer Bilimler Enstitüsü, 2018) Aktas Ozulus, Ö.; Ariman, N.; Iscen, F.; Kose, U.; Sengun, B.; Ergun, S.; Other; Other
    The Coolant Boiling in Rod Arrays-Two Fluid (COBRA-TF) code is known as a best estimate code, used to perform thermal-hydraulic analyses for a light water reactor vessel. On the other hand, ZEBRA computer code is used to perform closed channel analysis, basically for educational purposes. In this study, the hottest channels of two Russian type nuclear reactors, namely VVER-1000 and VVER-1200, are modelled as one single channel by using COBRA-TF and ZEBRA thermal hydraulic codes. Both outputs of these two codes and operating conditions of hot channels of these different type reactors are compared. Clad outside temperatures, clad inside temperatures, fuel centreline temperatures, pressure drops, qualities and critical heat fluxes are main operating parameters that are compared. Besides, look-up table that is used to calculate critical heat flux for triangular nuclear fuel assembly pitch and outputs of COBRA-TF and ZEBRA codes are also compared. Some of the correction factors which are used on the look-up table results, if/when necessary are included for the calculations as well.