SUBCHANNEL THERMOHYDRAULIC CALCULATIONS OF REACTOR CORE FOR WWER-SCP 30 MW (TH) UNDER NOMINAL CONDITIONS

Authors & Affiliations

Kartashov K.V.
A.I. Leypunsky Institute for Physics and Power Engineering, Obninsk, Russia

Kartashov K.V. - Assistant Director General for Development RPS, Cand. Sci. (Tech.), A.I. Leypunsky Institute for Physics and Power Engineering. Contacts: 1, pl. Bondarenko, Obninsk, Kaluga reg, Russia, 249033. Tel. (484)399-81-83; e-mail: This email address is being protected from spambots. You need JavaScript enabled to view it.

Abstract

Results of a comprehensive neutron-physical and thermohydraulic calculation for the two-way core experimental WWER-SCP-30 are presented in the paper. Following types of the fuel are considered: uranium dioxide and depleted uranium (DU) enriched by plutonium obtained from spent fuel from WWER. At the first stage, neutronic characteristics of the core sector are determined with the program ACADEM. At the second stage, these data are used as inputs for the program SUP (supercritical pressure), designed for large-scale calculation of thermal-hydraulic characteristics of the reactor core with supercritical fluid. As a result, the averaged thermal hydraulic parameters over cross section subassembly are calculated. At the final stage, the program MIF-SCP performs a detailed analysis of thermal-hydraulic characteristics of the most loaded channels in subassembly. From results of neutronic and thermohydraulic calculations it is clear that at the same geometric characteristics of subassembly the number of subassemblies in the core with MOX fuel is less than with UO₂, respectively, also, the diameter of the core is less, too. By using MOX fuel we have more uniform distribution of coolant temperature at the core outlet. Maximum flow temperature in subassemblies with UO₂ peripheral zone is 440°C and 640°C for central zone, in subassemblies with MOX fuel for the peripheral zone – 430°C, and the central zone – 540°C.

Keywords
experimental reactor, core, neutron-physical calculation, thermohydraulic calculation, subchannel code, nominal, supercritical parameters of water, central zone, peripheral zone, the most loaded channels in subassembly

Article Text (PDF, in Russian)

References

UDC 536.24.621.039

Problems of Atomic Science and Technology. Series: Nuclear and Reactor Constants, 2016, issue 3, 3:12