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The paper presents the results of an experimental study to investigate the coolant interaction in adjoining fuel assemblies in the VVER reactor core composed of TVSA-T and upgraded TVSA FAs. The processes of the in-core coolant flow were simulated in a test wind tunnel. The experiments were conducted using models representing different portions of the VVER reactor core fuel bundle and consisted in measuring the radial and axial airflow velocities in representative areas within the FAs and in the interassembly space. The results of the experiments can be translated to the full-scale conditions of the coolant flow with the use of the fluid dynamics simulation theory. The measurements were performed using a five-channel pressure-tube probe. The coolant flow pattern in different portions of the fuel bundle is represented by distribution diagrams and distribution maps for the radial and axial velocity vector components in the representative areas of the models. An analysis for the spatial distribution of the radial and axial velocity vector components has made it possible to obtain a detailed pattern of the coolant flow about the FA spacer, mixing and combined spacer grids of different designs. The accumulated database for the coolant flow in FAs of different designs forms the basis for the engineering justification of the VVER reactor core reliability and serviceability. The investigation results for the coolant interaction in adjoining TVSA FAs of different designs have been adopted for the practical use at JSC Afrikantov OKBM to estimate the heat-engineering reliability of the VVER reactor cores and have been included in the database for verification of computational fluid dynamics (CFD) codes and detailed by-channel calculation codes.

High fidelity nuclear reactor fuel assembly simulation using CFD method is an effective way for the structure design and optimization. The validated models and user practice guidelines play critical roles in achieving reliable results in CFD simulations. In this paper, the international benchmark MATiS-H is studied carefully and the best user practice guideline is achieved for the rod bundles simulation. Then a 5 × 5 rod bundles model in the advanced pressurized water reactor (PWR) is established and the detailed three-dimensional thermal-hydraulic characteristics are investigated. The influence of spacer grids and mixing vanes on the flow and hear transfer in rod bundles is revealed. As the coolant flows through the spacer grids and mixing vanes in the rod bundles, the drastic lateral flow would be induced and the pressure drop increases significantly. In addition, the heat transfer is enhanced remarkably due to the strong mixing effects. The calculation results could provide meaningful guidelines for the design of advanced PWR fuel assembly.

For the validation of advanced two-fluid CFD models at boiling water reactor (BWR) conditions, it is necessary to have good experimental data challenging for the code by reflecting relevant flow phenomena. The present experimental study aims to provide detailed film thickness distributions in a BWR-like subchannel geometry, including spacer grids and heat transfer. This is achieved using chloroform as a working fluid, indirectly heated by hot water channels simulating BWR fuel rods. X-ray equipment was used to produce tomographic 3D-data of time-averaged attenuation coefficients which essentially represent the liquid holdup. The conversion to film thickness was done by integrating the attenuation coefficient along lines perpendicular to the rod wall and using an effective chloroform attenuation coefficient determined by numerical simulation of the experimental setup, including X-ray sensor behavior. The simulation method was validated by using two different sensors types. Measurements were performed over the whole cross-section of the channel and an axial length of 11 cm. The experiments cover a range of liquid flow rates in the pre- and post-dryout regime., Nuclear Engineering and Design, 380, ISSN:0029-5493, ISSN:1872-759X

International audience; The vibration of rods is of prime importance to estimate their wearing due to the frictions with grids' elements. To predict this wearing, the forces exerted on the rods need to be identified and quantified. Pressure fluctuations are explored using an instrumented rod centred within a 5x5 rods bundle. Two configurations of spacer-grids are used: with and without mixing vanes respectively noted WMV and NMV. The position of the measurement point is rotated over 360° and its distance from the grids is varied from 0.5D$_h$ to 20D$_h$. The results show the non-homogeneity of the azimuthal distribution of the pressure fluctuations in the near wake of the spacer-grids for both configurations. The return to a homogeneous azimuthal distribution of the pressure fluctuations is faster without mixing vanes than with mixing vanes. The intensities of the pressure fluctuations are (in average) higher for WMV than for NMV: higher by 13% and 136% respectively at distances 0.5D$_h$ and 20D$_h$ from the grids. Power law approximations show decreases of $\langle$$P'_{rms}$$\rangle$$_\theta$ like $(y/D_h)$$^{-0.46}$ and $(y/D_h)$$^{-0.68}$. respectively for the configurations WMV and NMV. The profiles of mean velocity and velocity fluctuations are explored around the instrumented rod using LDV. Results show oscillation of the profiles and a damping with the distance from the grids. Power law approximations show decreases of $\langle$$V'_{rms}$$\rangle$$_{Psq}/V_{flow}$ ⁄ like $(y/D_h)$$^{-0.38}$ and $(y/D_h)$$^{-0.46}$. respectively for WMV and NMV. Between 2D$_h$ and 15D$_h$, the dimensionless ratio of "pressure fluctuations to velocity fluctuations" $\langle$$P'_{rms}$$\rangle$$_\theta$/$1\over 2$$\rho$ $\langle$$V'_{rms}$$^2$$\rangle$$_{Psq}$ is found to increase by 70% and 100% respectively for the configurations NMV and WMV. This increase demonstrated by the combination of pressure and velocity measurements further reveals the complexity of the turbulent flow in the wake of spacer-grids.

Spacer grid assembly supporting the nuclear fuel rods is an important concern in the design of structural components of a Pressurized Water Reactor (PWR). The spacer grid is composed by springs and dimples which are formed from a strip sheet by means of blanking and stamping processes. In this paper, the blanking process and tooling parameters are evaluated by means of a 2D plane-strain finite element model in order to evaluate the punch load and quality of the sheared edges of Inconel 718 strips used for nuclear spacer grids. A 3D finite element model is also proposed to predict the tooling loads resulting from the stamping process of a preformed Inconel 718 strip and to analyse the residual stress effects upon the spring and dimple design geometries of a nuclear spacer grid., {"references":["J. Wang, N. Kim, and H. Lee, \"Ductile fracture in the shearing process\nof Zircaloy sheet for nuclear fuel spacer grids\". In Metal and Materials\nInternational, vol. 18, No. 2, 2012, pp. 303-316.","C. Poizat, L. Champagne, L. Daridon, L. Ahzi, C. Husson, and L. Merle,\n\"Modeling and simulation of thin sheet blanking using damage and\nrupture criteria\". In International Journal of Forming Process, vol. 8,\nNo.1, 2005, pp.29-47.","A. L. Gurson, \"Continuum theory of ductile rupture by void nucleation\nand growth. I. Yield criteria and flow rules for porous ductile media\". In\nJournal of Engineering Materials and Technology. vol. 99, No. 1, 1977,\npp. 2-15.","A. Needleman and V. Tvergaard, \"An analysis of ductile rupture in\nnotched bars\". In Journal of the Mechanics and Physics of Solids.\nvol. 32, No 6, 1984, pp. 461-490.","ABAQUS user's manual, Version 6.9, Dassault Systemes Simulia Corp.,\nProvidence, RI, USA, 2009."]}

Fuel elements of PWR type nuclear reactors consist of rod bundles, arranged in a square array, and held by spacer grids. The coolant flows, mainly, axially along the rods. Although such elements are laterally open, experiments are performed in closed type test sections, originating the appearance of subchannels with different geometries. In the present work, utilizing a test section of two bundles of 4x4 pins each, experiments were performed to determine the friction and the grid drag coefficients for the different subchannels and to observe the effect of the grids in the crossflow, in cases of inlet flow maldistribution.

The paper deals with the modelling and modal analysis of the large package of identical parallel rods linked by transverse springs (spacer grids) placed on several level spacings. The rod discretization by finite element method is based on Rayleigh beam theory. For the cyclic and central symmetric package of rods (such as fuel rods in nuclear fuel assembly) the system decomposition on the identical revolved rod segments was applied. A modal synthesis method with condensation is used for modelling of the whole system. The presented method is the first step for modelling the nuclear fuel assembly vibration caused by excitation determined by the support plate motion of the reactor core.