DEVELOPMENT AND ADAPTATION OF VORTEX REALIZABLE MEASUREMENT SYSTEM FOR BENCHMARK TEST WITH LARGE SCALE MODEL OF NUCLEAR REACTOR

The last decades development of applied calculation methods of nuclear reactor thermal and hydraulic processes are marked by the rapid growth of the High Performance Computing (HPC), which contribute to the active introduction of Computational Fluid Dynamics (CFD). The use of such programs to justify technical and economic parameters and especially the safety of nuclear reactors requires comprehensive verification of mathematical models and CFD programs. The aim of the work was the development and adaptation of a measuring system having the characteristics necessary for its application in the verification test (experimental) facility. It’s main objective is to study the processes of coolant flow mixing with different physical properties (for example, the concentration of dissolved impurities) inside a large-scale reactor model. The basic method used for registration of the spatial concentration field in the mixing area is the method of spatial conductometry. I

n the course of the work, a measurement complex, including spatial conductometric sensors, a system of secondary converters and software, was created. Methods of calibration and normalization of measurement results are developed. Averaged concentration fields, nonstationary realizations of the measured local conductivity were obtained during the first experimental series, spectral and statistical analysis of the realizations were carried out.

The acquired data are compared with pretest CFD-calculations performed in the ANSYS CFX program. A joint analysis of the obtained results made it possible to identify the main regularities of the process under study, and to demonstrate the capabilities of the designed measuring system to receive the experimental data of the «CFD-quality» required for verification.

The carried out adaptation of spatial sensors allows to conduct a more extensive program of experimental tests, on the basis of which a databank and necessary generalizations will be created. The received information allows to answer a number of questions related to scaling up the results of CFD calculations at the fullscale parameters of pressurized nuclear reactors. 

1. Review of Uncertainty Methods for Computational Fluid Dynamics Application to Nuclear Reactor Thermal Hydraulics. OECD, NEA, 2016, 132 c.

2. Rohde U., Kliem S., Hohne T., Karlsson R., Hemstrom B., Lillington J., Topipila T., Elter J., Bezrukov Y. Fluid mixing and flow distribution in the reactor circuit, measurement data base. Nuclear Engineering and Design, 2005, vol. 235, pp. 421–443. doi: 10.1016/j.nucengdes.2004.08.045

3. Farkas I., Hutly E., Farkas T., Takaxs A., Guba A., Toth I. Validation of computational fluid dynamics calculation using Rossendorf coolant mixing model flow meaurements in primary loop of coolant in a pressurized water reactor model. Nuclear engineering and technology, 2016, vol.48, pp. 941–951. doi: 10.1016/j.net.2016.02.017

4. Kim K., Euh D.J., Kwon T.S. Schleicher E. Mixing behaviors in a downcomer and core inlet. Proceedings of Specialists Workshop on Advanced instrumentation and Measurement Techniques for Nuclear Reactor Thermal Hydraulics (SWINTH-2016), 2016, paper #016, pp. 1–14.

5. Kim K., Euh D.J., Chu I.C., Youn Y.J., Choi H.S., Kwon T.S. Experimental study of the APR+ reactor core flow and pressure distributions under 4-pump running conditions. Nuclear Engineering and Design, 2013, vol. 265, pp. 957–966. doi: 10.1016/j.nucengdes.2013.07.021

6. Forgione N., Baudanza V., Angelo I., Martelli D. Downcomer and lower plenum experimental facility. Report RdS, ENEA,UNIPI, 2011.

7. Volkov V. Golibrodo L., Krutikov A. Kudryavtsev O., Nadinsky U., Skibin A. [Verification of CFD reactor model on the fast borated water injection system test results]. Materialy 10-j MNTK «Obespechenie bezopasnosti AJeS s VVER», OKB «GIDROPRESS» [Proceedings of the 10th international conference «NPP with VVER safety assurance» JSC OKB Gidropress], 2017 (in Russian).

8. Dmitriev S.M., Khrobostov A.E., Barinov A.A., Glavny V.G. [Application features of spatial conductometry sensors in modelling of coolant flow mixing in nuclear power unit equipment]. Devices and Methods of Measurements, 2016, vol. 7, no. 3, рp. 247–255 (in Russian).

9. daSilva M.J., Eduardo N.S., Vendruscolo T.P. High speed multichannel impedance measuring system. Acta IMEKO, 2012, vol. 1, pp. 36–41.

10. Budnikov A.V., Sveshnikov D.N., Romanov R.I. [Problems of CFD-program scaling for mixing processes description in nuclear power units]. Materialy conferencii «Problemy verifikacii i primenenya CFD-kodov v atomnoy energetike» [Problems of CFD-codes verification and application in atomic energy], 2016 (in Russian)