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This work is an experimental study into the hydrodynamics of the coolant flow in the in-vessel pressure duct of a pressurized water reactor. In the article, an experimental testbench and a nuclear reactor model under investigation are described, the measurement methods are set forth, and the operating variables at which the study was conducted and the obtained results are provided. In the experiments, the mixing of the loop coolant flows inside the model of the nuclear reactor downcomer was simulated. The study was conducted on the high-pressure aerodynamic testbench of Alekseev State Technical University in Nizhny Novgorod. The scale model of the nuclear reactor had the structural components characteristic of loop-type reactor units, such as the annular downcomer and the bottom pressure vessel. The experiments were conducted at Reynolds numbers within the 20 000–50 000 range measured in the annular gap of the downcomer of the model. The axial velocity field at the inlet to the reactor core simulator was investigated using a pneumometric probe. The temperature field was recorded in the experiment by the impurity diffusion method, i.e., by introducing of a contrast tracer into one of the loops of the model. The degree of mixing the flows was estimated by the admixture (tracer) concentration at the inlet to the core simulator. Propane was used as the contrast admixture. The study has yielded the spatial distribution of the tracer in the coolant flow in the annular downcomer and in the bottom pressure vessel. The data on the distribution of the contrast admixture are presented in the form of charts. The swirling of the coolant flow in the in-vessel pressure duct has been analyzed. It has been shown that the mixing intensity in the bottom pressure vessel is affected by the central vortex with the central axis. The parameters of mixing the admixture in the model of the in-vessel pressure duct have been estimated.

The results from studying the coolant flow in fuel assemblies downstream of mixing spacer grids of a PWR reactor are presented. The aim of the study is to estimate the effect from using different designs of mixing grids. For this purpose, experimental investigations were carried out on an aerodynamic test facility with scaled models of fuel assembly fragments containing mixing spacer grids of different designs. Two adjacent cells of the guiding channel zone were selected as the representative study area. The specific structural feature of these cells is that the turbulizers installed on the intensifying grids have different spatial orientation. The general flow pattern is represented by the vector fields of tangential velocities and by the distribution of relative velocities in the gaps between the fuel rods and the guiding channel. For estimating the effect from using the grid designs with respect to coolant flow mixing, the intracell vortex formation coefficients and intercell mixing coefficients were analyzed. It has been shown from the performed analysis that, owing to the use of the alternative design with a pair of cell deflectors oriented in a changed direction, the intercell transfer coefficient and the intracell vortex formation coefficient were increased by a factor of 1.13 and 2.2, respectively, in comparison with those in the basic design. Hence, the use of the grid of the alternative design is more preferable for obtaining better mixing of the coolant flow. The accumulated database on coolant flow in the Square FA serves for engineering substantiation of the PWR reactor cores. The results from experimental studies are used for verification of CFD codes developed both in Russia and abroad and of programs for detailed cell-wise analysis of the cores for reducing the conservatism in substantiating their thermal reliability.

The article presents the results from experimental investigations into local coolant flow fluid dynamics that were carried out on the fragmental model of a VVER-type pressurized water reactor’s mixed core consisting of two types of fuel assemblies (FAs): TVSA-T (one segment) and TVSA-T.mod.2 (two segments). The coolant flow processes in the fuel rod bundle were simulated on an aerodynamic test bench. The pressure field in the flow was measured using a five-channel pneumometric probe. The obtained pressure field in the flow was recalculated for the coolant velocity vector according to the dependences derived during calibration. For drawing up a detailed flow motion pattern, a characteristic model cross section area was separated, which included the space between the assemblies and four fuel rod rows in each TVSA-type fuel assembly. The spatial distribution of coolant flow velocity projections was analyzed, as a result of which it became possible to reveal the regularities associated with the streamlining of spacer-, mixing-, and combined-spacer grids in TVSA assemblies by coolant; to determine coolant cross flows resulting from streamlining of hydraulically nonidentical grids and determine their localization in the experimental model longitudinal and cross sections; and to reveal the effect of accumulating flow hydrodynamic disturbances in the model longitudinal and cross sections resulting from the staggered arrangement of hydraulically nonidentical grids. The results obtained from investigations of inter-assembly interaction of coolant between the neighboring TVSA-T and TVSA-T.mod.2 assemblies have been adopted for practical use at AO Afrikantov OKBM in estimating the thermal reliability of VVER-type reactor cores and have been included in the database for verification of computation fluid dynamics computer programs (CFD codes) and detailed cell-wise numerical analysis of the core of VVER reactors.

The experimental investigation of local hydrodynamic characteristics of the coolant flow downstream of a mixing spacer grid (MSG) in characteristic regions of a Kvadrat fuel assembly in PWRs was performed. The importance of this investigation is due to the fact that MSGs of interest can enhance heat-and-mass transfer. Finding the best design of the grid is required for corroboration of the thermal engineering reliability and operability of the PWR core. The paper presents a description of the test facility and a model of a fuel assembly fragment, the investigation procedure, and discussion of the results. The experiments were performed in the aerodynamic test facility on a scaled model of a fragment of a Kvadrat fuel assembly by simulating the water coolant flow using an air flow according to the hydrodynamic similarity theory. Within the scope of the investigation, the transverse velocity fields in the coolant flow were studied in the characteristic cross-sections of the fuel assembly. The transverse velocity distribution was measured with a five-channel pneumometric probe that can determine the module and direction of a flow velocity vector at an investigated point. The experimental results are presented in the form of distribution of relative transverse velocity downstream of MSG in the characteristic regions of the fuel assembly, such as standard cells, guide channel, and a gap between fuel rods. Based on the experimental data, the coolant flow features have been revealed, and the regularities in the development of transverse velocity field downstream of the mixing spacer grid under investigation have been determined. These experimental data are required for assessment of the mixing spacer grid effectiveness, verification of 3D CFD-programs, and applicable cell-by-cell codes for calculation of a PWR core with a Kvadrat fuel assembly.

Results of research works on studying local hydrodynamics and mass transfer for coolant flow in the characteristic zones of PWR reactor fuel assemblies in case of using belts of mixing spacer grids are presented. The investigations were carried out on an aerodynamic rig using the admixture diffusion method (the tracer-gas method). Certain specific features pertinent to coolant flow in the fuel rod bundles of Kvadrat-type fuel assemblies were revealed during the experiments. The obtained study results were included in the database for verifying computation fluid dynamics computer codes and detailed cell-wise calculations of reactor cores with Kvadrat-type fuel assemblies. The obtained results can also be used for more exact determination of local coolant flow hydrodynamic and mass transfer characteristics in assessing thermal reliability of PWR reactor cores.