Izvestiya vuzov. Yadernaya Energetika

The peer-reviewed scientific and technology journal. ISSN: 0204-3327

Development of a Methodological Approach for Analytical Study of the Coolant Flow in the Process of Sodium-Cooled Reactor Cooldown

11/19/2020 2020 - #04 Thermal physics and thermal hydraulics

Didenko D.V. Baluyev D.E. Nikanorov O.L. Rogozhkin S.A. Shepelev S.F. Aksenov A.A. Zhestkov M.N. Shchelyaev A.E.

DOI: https://doi.org/10.26583/npe.2020.4.07

UDC: 621.039.5:(532+536) БН+621.039.513:621.039.526

A methodological approach for an analytical study of thermohydraulic processes occurring in a fast sodium cooled reactor was developed using the domestic software package for computational fluid dynamics – FlowVision. This approach considers an integral layout of the primary equipment of the reactor plant, peculiarities of heat exchange in a liquid metal coolant, and also makes it possible, using proved simplifications, to simulate heat-mass-exchange in the course of coolant flow through the core and heat exchanging equipment of the reactor. In particular, the methodological approach can be used to justify safety during reactor cooldown and for other analytical studies that require simulating the core and heat exchange equipment of an integral-type reactor.

A brief overview was given of the methodological approaches developed previously to study cooldown processes of liquid metal cooled reactors. The general principles of these approaches, their advantages and disadvantages were highlighted.

A 3-D analytical model consisting of one loop (one-fourth part of the reactor) of an advanced reactor was developed. The authors justified the applicability of the FlowVision gap model for simulating the space between the heat-generating assemblies of the core (inter-package space) as well as the porous skeleton model for simulating the heat exchange equipment of the reactor. An analytical study of the nominal flow regime of the coolant in the reactor was carried out. The paper shows that the developed methodological approach is applicable to solving problems of coolant flow in various operating modes of liquid metal cooled reactor plants.

References

  1. Pialla D., Tenchine D., Li S., Gauthe P., Vasile A., Baviere R., Tauveron N., Perdu F., Maas L., Cocheme F., Huber K., Cheng X. Overview of the system alone and system/CFD coupled calculations of the PHENIX Natural Circulation Test within the THINS project. Nuclear Engineering and Design. 2015, v. 290, pp. 78-86.
  2. Rakhi, Anil Kumar Sharma, Velusamy K. Integrated CFD investigation of heat transfer enhancement using multi-tray core catcher in SFR. Annals of Nuclear Energy. 2017, v. 104, pp. 256-266.
  3. Merzari E., Shemon E., Yu Y., Thomas J., Obabko A., Jain R., Mahadevan V., Solberg J., Ferencz R., Whitesides R. Full Core Multi-Physics Simulation with Offline Core Deformation. ANL/NE-15/42, Nuclear Engineering Division, Argonne National Laboratory, 2015.
  4. Korsun A.S., Merinov I.G., Kharitonov V.S., Bayaskhalanov M.V., Chudanov V.V., Aksenova A.E., Prvichko V.A. Numerical Simulation of Thermal-Hydralic Processes Liquid-Metal Cooled Fuel Assemblies in the Anisotropic Porous Body Approximation. Teploenergetika. 2019, no. 4, pp. 12-22 (in Russian).
  5. International Atomic Energy Agency, Status of Fast Reactor Research and Technology Development, IAEA-TECDOC-1691. Vienna, IAEA, 2012, 832 p.
  6. Park Jong-Pil, Jeong Ji Hwan, Lee Tae-Ho. Scientific design of a large-scale sodium thermal-hydraulic test facility for KALIMER – Part II: Validation of reactor pool design using CFD analyses. Annals of Nuclear Energy. 2015, v. 76, pp. 439-450.
  7. Rogozhkin S.A., Aksenov A.A., Zhluktov S.V., Osipov S.L., Sazonova M.L., Fadeev I.D., Shepelev S.F., Shmelev V.V. Development and verification of a turbulent heat transport model for sodium-based liquid metal coolants. Vychislitel‘naya Mekhanika Sploshnykh Sred. 2014, v. 7, no. 3, pp. 306-316 (in Russian).
  8. Rogozhkin S.A., Fadeev I.D., Shepelev S.F., Aksenov A.A., Mosunova N.A., Frick P.G. V&V Status of CFD Codes Applied to BN Reactors. Proc. of the International Conference on Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable Development (FR-17). Paper IAEA-CN245-418. Yekaterinburg, Russian Federation, 2017, 10 p.
  9. Attestation Passport of the Computer Program FlowVision. Registration No 492 from 19.12.2019 (in Russian).
  10. Ozturk U., Soganci S., Akimov V., Tutkun M., Aksenov A. Validation of FlowVision CFD on ICCS2015 Test Case: Application of Gap Model and SGGR for Leakage Flow Prediction in a Dry Screw Compressor. IOP Conference Series: Materials Science and Engineering. 2019, v. 604 (012010).
  11. Rogozhkin S.A., Aksenov A.A., Pakholkov V.V.,Zhluktov S.V., Zhestkov M.N., Shepelev S.F., Shmelev V.V. Development of methodology for computational analysis of thermos-hydraulic processes in fast-neutron reactor with FlowVision CFD software. Komp‘yuternye Issledovaniya i Modelirovanie. 2017, v. 9, no. 1, pp. 87-94 (in Russian).
  12. User’s Guide FlowVision. Version 3.12.01. Moscow. «TESIS» Ltd. Publ., 2020, 1549 p. (in Russian).

liquid metal cooled reactor integral layout methodological approach CFD FlowVision core LMS reactor cooldown inter-wrapper space

Link for citing the article: Didenko D.V., Baluyev D.E., Nikanorov O.L., Rogozhkin S.A., Shepelev S.F., Aksenov A.A., Zhestkov M.N., Shchelyaev A.E. Development of a Methodological Approach for Analytical Study of the Coolant Flow in the Process of Sodium-Cooled Reactor Cooldown. Izvestiya vuzov. Yadernaya Energetika. 2020, no. 4, pp. 74-85; DOI: https://doi.org/10.26583/npe.2020.4.07 (in Russian).