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[en] Three concepts of reactor plant (RP) with supercritical water coolant VVER-SCP (SCWR), which refer to the Generation-4 systems, were developed in Russia: — VVER-SCP-1700 - one-circuit high-power RP, with fast-resonant neutron spectrum in the core; — SPCS-600 - two-circuit RP, with pseudo-vapor supercritical primary coolant and fast neutron spectrum in the core; — VVER-SCP-I - two-circuit RP, with supercritical water coolant in the primary circuit and regulated neutron spectrum. Up to the moment, OKB ^GIDROPRESS implements engineering development of VVER-SCP-1700 and performs thermal-hydraulic analyses. Several problems, which were discovered during the subchannel calculations of the coolant flow in FA of one-circuit VVER-SCP-1700 must be studied with the first priority. The report emphasizes the problem of the displacement of coolant from the more heatstressed part of the bundle to the less heat-stressed. The results of the calculations concerning this problem are presented. New specification of the future thermal-hydraulic subchannel calculations in the framework of development and verification of TEMPA-SC code is presented.

[en] At present OKB 'GIDROPRESS' (Podolsk) and IPPE (Obninsk) are engaged in performing the conceptual design and computational studies of the single-loop reactor plant VVER-SCP cooled by supercritical water. Two variants of coolant flow arrangement in the core are considered: single-pass and double-pass. The paper emphasizes the thermohydraulic features of VVER-SCP single-pass core. The computational studies performed for VVER-SCP core have revealed the necessity of the following: - throttling of the fuel assembly inlet for the coolant flow thermohydraulic stability; - hydro-profiling of the core for equalizing the coolant temperatures at the outlets of fuel assembles with different power; - limitation of power peaking in fuel rods in the fuel assembly. The paper gives a brief description of VVER-SCP design and the results of the computational studies and principal engineering solutions, aimed at assurance of coolant flow stability and equalization of coolant heating in different fuel assemblies, are presented. A brief description of the applied methods and codes is given. (author)

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[en] The report provides an overview of the experiments conducted earlier in OKB; ‘GIDROPRESS’ to study the conditions of fuel assemblies cooling in loss-of-coolant accidents. A modernized experimental facility ‘Reflooding stand” is described and the results of new experiments are presented. In OKB ‘GIDROPRESS’ several experimental facilities were made to study the conditions of fuel assemblies cooling in loss-of-coolant accidents. The first was an integral double-loop test facility (safety stand), structurally simulating the first reactor circuit. It was used to study the processes characteristic of instantaneous break of main circulation piping with total cross section. The initial task of the study was to obtain experimental data to justify the computational methods for analyzing the safety of the designed NPP Loviisa. After performing the studies relating to reactor VVER-440, the work was continued relating to reactor VVER-1000. In the following, the experimental facility ‘Reflooding stand” was built (1976), which schematically simulates the first circuit of VVER-440 with a full-scale fuel assembly model of 126; fuel rod simulators as a core model. In 2003, the experimental facility was upgraded to study the cooling of VVER-1000 fuel assembly. In 2015, there was performed a series of studies on the effect of mixing grids on the cooling of TVS-2M fuel assembly in an accident with a break of the main circulation piping. In 2018 to obtain additional data for verification of calculation codes, a new modernization of the experimental facility was completed, during which its hydraulic scheme was significantly modified, that expands the possibilities of simulating various accident scenarios. The control of valves and pumps was also automated. Monitoring instrumentation was significantly improved. As fuel assembly model, a 127-rod bundle enclosed in a hexahedral cladding for key size of 147 mm was installed. The bundle consists of 120 heated fuel rod simulators placed on a triangular grid with a pitch of 12.75 mm in spacer grids located at distances corresponding to full-scale fuel assembly of NPP-2006. A central support tube is mounted into the center of the bundle, and six simulators of guide channels, 13 mm in outer diameter, are installed along the third row from the center. The model of the fuel assembly head was improved, and more closely corresponds to the design of the standard fuel assembly head from the point of view of organizing steam counterflows and cooling water for the regimes of flooding from above. A part of the planned experiments with the implementation of various methods of water supply was performed: from below with direct water supply under fuel assembly model; from below with water supply to the upper part of the reactor downcomer model; from above with water supply to the model of the reactor collecting chamber combined supply methods. It was concluded about the effect of the flooding method on cooling, and the obtained data were compared with the results of the studies obtained in other test facilities. Keywords: VVER FA, loss-of-coolant accident, cladding temperature of fuel rod simulators, experimental studies

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