Thermal-Hydraulic Characteristics of a Conceptual Reactor Plant with a Fast-Neutron Nuclear Reactor Cooled by the Carbon Dioxide

Abstract

The paper considers cooling conditions of a conceptual fast-neutron nuclear reactor that is using carbon dioxide as the coolant at low pressure in the primary circuit. The concept peculiarity lies in the close-to-atmospheric pressure of the primary circuit coolant and the reduced fuel volumetric energy release in the active zone. The reactor plant under consideration is proposed for implementation with a serial turboplant for thermal power engineering with the supercritical parameters of the K-800-240 working fluid. Reactor geometric parameters and the primary circuit coolant parameters are computed. They are ensuring fuel cooling both in the nominal operating mode and in the transient modes associated with the loss of power supply for the own needs and the loss of the main coolant with its replacement for the atmospheric air. The paper considers two transient scenarios, including removal of the residual energy releases and the reactor operation at the reduced power level, to assess a possibility of using the emergency cooldown system as the standard system in heating the reactor to the nominal parameters. Using the TPP and Desna software, a power unit model with installation of the reactor under study is developed making it possible to compute the non-stationary transient modes

Please cite this article in English as:

Semishin V.V., Kavun O.Yu. Thermal-hydraulic characteristics of a conceptual reactor plant with a fast-neutron nuclear reactor cooled by the carbon dioxide. Herald of the Bauman Moscow State Technical University, Series Mechanical Engineering, 2025, no. 2 (153), pp. 118--135 (in Russ.). EDN: PPDHGZ

References

[1] Kavun O.Yu., Lifshits A.M., Semishin V.V. Concept gas-cooled internal safe nuclear reactor on fast neutrons. Avtomatizatsiya i IT v energetike, 2021, no. 4, pp. 12--23 (in Russ.).EDN: JIVXLO

[2] Zolotarev A.V., Korobkova P.A., Semishin V.V. [Development of basic design solutions for gas-cooled fast neutron nuclear reactor]. Studencheskaya nauchnaya vesna. Sb. tez. dokl. Vseros. stud. konf. [Student’s Spring. Abs. Russ. Student Conf.] Moscow, Nauchnaya biblioteka Publ., 2022, pp. 53--54 (in Russ.). EDN: LDGAOQ

[3] Kalincheva T.V., Semishin V.V. [Investigation of neutron-physical characteristics and fuel composition of gas-cooled fast reactor.]. Studencheskaya nauchnaya vesna. Sb. tez. dokl. Vseros. stud. konf. [Student’s Spring. Abs. Russ. Student Conf]. Moscow, Nauchnaya biblioteka Publ., 2022, pp. 55--56 (in Russ.). EDN: CMVJON

[4] Bagdasarov Yu.E., Poplavskiy V.M., Matveev V.I., et al. Natural safety of advanced nuclear technology based on BN-800. Atomnaya energiya, 2001, vol. 90, no. 6, pp. 438--444 (in Russ.).

[5] Ponomarev-Stepnoy N.N. Two-component nuclear power system with a closed nuclear fuel cycle on the basis of BN and VVER. Atomnaya energiya, 2016, vol. 120, no. 4, pp. 183--191 (in Russ.).

[6] Muravyev E.V. Relevance of closed nuclear fuel cycle. Atomnaya energiya, 2011, vol. 111, no. 6, pp. 334--342 (in Russ.).

[7] Darilek P., Zajac R. ALLEGRO --- introduction to GFR. Proc. Twenty-First Symposium of Atomic Energy Research on WWER Physics and Reactor Safety. Budapest, IAEA--Kiadja and KFKI Atomenergia Kutatointezet, 2011, pp. 1--10.

[8] Hatala B. Gas cooled fast reactor. 13th GIF-IAEA Interface Meeting Presentations. Vienna, IAEA, 2019.

[9] Phenix and Super-Phenix reactors. In: Liquid metal cooled reactors: experience in design and operation. Vienna, IAEA, 2007, pp. 57--101.

[10] Dementyev B.A. Yadernye energeticheskie reaktory [Nuclear power reactors]. Moscow, Energoatomizdat Publ., 1990.

[11] Glazov A.G., Leonov V.N., Orlov V.V., et al. BREST reactor and the stationary nuclear fuel cycle. Atomnaya energiya, 2011, vol. 103, no. 1, pp. 15--21 (in Russ.).

[12] Semishin V.V., Kavun O.Yu. [Calculation Study of Thermal Hydraulic Characteristics of the Fast Neutron Nuclear Reactor Cooled by CO2]. Teplofizika -- 2022. Sb. tez. dokl. nauch.-tekh. konf. [Thermophysics-2022. Abs. Sc.-Tech. Conf.]. Obninsk, GNTs RF--FEI Publ., 2022.

[13] Volchkov V.I., Volfovskiy S.A., Kovalev I.A., et al. Parovye turbiny sverkhkriticheskikh parametrov LMZ [Steam Turbines of Supercritical Parameters of LMZ]. Moscow, Energoatomizdat Publ., 1991.

[14] Kavun O.Yu., Kuno M.Ya., Feyman V.G. [TPP software for thermal-hydraulic calculation of complex thermal-hydraulic networks]. Algoritmy i programmy dlya neytronno-fizicheskikh raschetov yadernykh reaktorov. Neytronika--98. Sb. tr. seminara MAE RF [Algorithms and Programs for Neutron-Physical Calculations of Nuclear Reactors. Neutronics--98. Proc. Seminar MAE RF]. Obninsk, FEI Publ., 1998, pp. 111--118 (in Russ.).

[15] Kavun O.Yu. Methodology of NPP power unit dynamics modeling implemented in the software complex "RADUGA-EU". Voprosy atomnoy nauki i tekhniki. Ser. Fizika yadernykh reaktorov, 1999, no. 2, pp. 17--39 (in Russ.).

[16] Kavun O.Yu., Polikarpova A.M., Pipchenko G.R. Development of models for the rapid assessment of critical safety functions status of NPP with VVER-type reactors. Yadernaya i radiatsionnaya bezopasnost [Nuclear and Radiation Safety Journal], 2018, no. 1, pp. 10--19 (in Russ.). EDN: XTXZHV

[17] Semishin V.V., Kavun V.O., Kavun O.Yu., et al. [Development of a model for express assessment of NPP power unit condition with BN-800 type reactor unit]. Innovatsii v atomnoy energetike. Tr. VI Ros. nauch.-tekh. konf. molodykh spetsialistov [Innovations in Nuclear Power Engineering. Proc. VI Ros. Sc.-Tech. Conf. of Young Specialists]. Moscow, AO "NIKIET" Publ., 2019, pp. 553--562 (in Russ.).

[18] Kartashov B.A., Kozlov O.S., Vinnikov I.K., et al. Modeling of technical system in the environment of program complex "MVTU". Vestnik DGTU [Vestnik of Don State Technical University], 2009, vol. 9, no. 4, pp. 643--647 (in Russ.).

[19] Anwyl С., Boxall С., Wilbraham R., et al. Corrosion of AGR fuel pin steel under conditions relevant to permanent disposal. Procedia Chem., 2016, vol. 21, pp. 247--254. DOI: https://doi.org/10.1016/j.proche.2016.10.035

[20] Pexton A.F. MAGNOX reactors and advanced gas-cooled reactors. Technical Committee Meeting on Gas-Cooled Reactors and Their Applications, 1986, pp. 11--28.

[21] Nonbol E. Description of the advanced gas cooled type of reactor (AGR). Roskilde, Riso National Laboratory, 1996.

[22] Bebjak S., Kvizda B. Gas-cooled fast reactor --- ALLEGRO decay heat removal studies. Proc. 23 Inter. Conf. on Applied Physics of Condensed Matter. Bratislava, University of Technology, 2017, pp. 29--33.

[23] Cheng L.Y., Wei T.-Y.-C. Decay heat removal in GEN IV gas-cooled fast reactors. Sc. Technol. Nucl. Install., 2009, vol. 2009, pp. 99--111. DOI: https://doi.org/10.1155/2009/797461