Izvestiya vuzov. Yadernaya Energetika

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

Investigation of the small break conditions in the primary circuit of a VVER-1000 reactor

9/20/2018 2018 - #03 Global safety, reliability and diagnostics of nuclear power installations

Belozerov V.I. Zhuk M.M.

DOI: https://doi.org/10.26583/npe.2018.3.12

UDC: 621.039.524.44

Modes with violation of the reactor facility cooling conditions in the VVER reactor primary circuit have been simulated using the TRAC-PD2 and Open FOAM thermohydraulic codes [1 – 3] based on energy and mass conservation equations for a three-dimensional unsteady flow of a two-phase mixture. Coupled simulation of the dynamics of neutronic and thermohydraulic processes [1 – 8] aims to improve the qualitative understanding and the quantitative notion of their effects on safety.

Studying these modes using the above thermohydraulic codes makes it possible to analyze the course of transients and certain emergency processes without using the industrial testing method, this providing the basis for solving the problems of ensuring the reliability, operational safety and efficiency of nuclear power plants.

A modern nuclear reactor is a complex system for studying and calculating which it is not enough to use simple theoretical models. Thermohydraulic calculations are an essential component of most engineering and technological development works in nuclear power. Since, in conditions of an NPP, it is not possible to use a technologically conventional way to verify and update the results and findings of an a priori analysis on the basis of industrial tests, investigations based on codes are used in some cases as the tools for studying and predicting the parameters of thermohydraulic processes in the reactor’s circulation circuit.

The main purpose of the study is to calculate and investigate, with the use of codes, modes with violation of the reactor facility cooling conditions in the primary circuit of a VVER reactor in order to determine the conformity of the calculated parameters to the acceptance criteria established by regulatory documentation.

References

  1. TRAC-PD2. An advanced best-estimated computer program for pressurized water reactor loss-ofcoolant accident analysis. NUREG/CR-2054, 1981.
  2. OpenFOAM User Guide Version 1.6. 2009.
  3. OpenFOAM Programmer’s Guide Version 1.6. 2009.
  4. Bolshagin S.N., Gorodkov S.S., Dementyev V.G., Oleynik D.S. Calculation of the VVER-1000 core neutron-physical characteristics by the Monte-Carlo method according to the MCU-PD program, and comparison of the results with the BIPR-7A, RADAR programs and experimental data, Materials of the XX-th seminar «NEITRONICA». Obninsk, October 27 – October 30, 2009. Obninsk. GNC RF-FEI Publ., 2009 (in Russian).
  5. Galanin A.D. Theory of a Heterogeneous Reactor. Moscow. Atomizdat Publ., 1971, 248 p. (in Russian).
  6. Weinberg A., Wigner E. The Physical Theory of Neutron Chain Reactors. Moscow. Inostrannaya Literatura Publ., 1961, 733 p. (in Russian).
  7. Ovchinnikov F.Ya., Semenov V.V. Operational regimes of water5water power reactors. Moscow. Energoatomizdat Publ., 1988, 359 p. (in Russian).
  8. MCNP – A General Monte Carlo N-Particle Transport Code, Version 5. X-5 Monte Carlo Team. Los Alamos National Laboratory Report LA-UR-03-1987.
  9. Safety of nuclear power plants. Handbook. Moscow. Concern «Rosenergoatom» Publ., 1994, 250 p. (in Russian).
  10. Reactor unit V5320. Technical description and safety information. Podol’sk. OKB «Gidropress» Publ., 1987, 62 p. (in Russian).
  11. NP-001-15. General provisions for ensuring the safety of nuclear power plants. Moscow. FBU «NTC YaRB» Publ., 57 p. (in Russian).
  12. Afrov A.M., Andrushechko S.A., Ukraintsev V.F., Vasilyev B.Yu., Kosourov K.B., Semchenkov Yu.M., Kokosadze E.L., Ivanov E.A. VVER51000: physical bases of operation, nuclear fuel, safety. Moscow. Universitetskay kniga, Logos Publ., 2006, 488 p. (in Russian).
  13. Kirillov P.L., Bogoslovskaya G.P. Heat and mass transfer in nuclear power plants. Moscow. Energoatomizdat Publ., 2000, 456 p. (in Russian).
  14. Kirillov P.L., Bobkov V.P., Zhukov A.V., Yuriev Yu.S. Handbook on thermohydraulic calculations in nuclear power. Vol. 1. Thermohydraulic processes in nuclear power plants. Ed. by P.L. Kirillov. Moscow. Izdat Publ., 2010, 776 p. (in Russian).
  15. Kuznetsov Yu.N. Heat exchange in the safety of nuclear reactors. Moscow. Energoatomizdat Publ., 1989, 296 p. (in Russian).
  16. Kutepov A.M., Sterman L.S., Styushin N.G. Hydrodynamics and heat transfer in steam generation. A manual for universities. Moscow. Vysshaya Shkola Publ., 1986, 448 p. (in Russian).
  17. Petrosyants A.M. Nuclear energy. Moscow. Nauka Publ., 1981, 272 p. (in Russian).
  18. Petukhov B.S., Polyakov A.F. Heat transfer in mixed turbulent convection. Мoscow. Nauka Publ., 1986, 192 p. (in Russian).
  19. Petukhov B.S. Heat transfer in a moving single5phase medium. Moscow. MEI Publ., 1993, 352 p. (in Russian).
  20. Galin N.M., Kirillov P.L. Heat and Mass Transfer (in Nuclear Power). Moscow. Energoatomizdat Publ., 1987, 375 p. (in Russian).
  21. Belozerov V.I., Sergeev V.V., Kazantsev A.A., Poznyakov A.N., Kanyshev M.Yu. Neutron-physical and thermohydraulic model of VVER-1000 for personnel training. Izvestia Vysshikh Uchebnykh Zavedenij. Yadernaya Energetika. 2008, no. 2, pp. 99-106 (in Russian).
  22. Kazantsev A.A., Sergeev V.V., Belozerov V.I., Efremov A.Yu. Modeling of transient processes for the VVER-1000 reactor. Izvestia Vysshikh Uchebnykh Zavedenij. Yadernaya Energetika. 2009, no. 1, pp. 98-104 (in Russian).
  23. Bolshov L., Strizhov V. SOCRAT – The System of Codes for Realistic Analysis of Severe Accidents. Proc. of ICAPP ’06 - Reno, NV USA, Paper 6439, 2006.
  24. Fuel Rod Behavior and Uncertainty Analysis by FRAPTRAN/TRACE/DAKOTA Code in Maanshan LBLOCA (NUREG/IA-0471).
  25. RELAP5/MOD3. Code Manual. Volume 1. Idaho Natural Engineering Laboratory, Idaho, 1995.
  26. LSDyna. Available at: http://www.lsdyna.ru/ (accessed: Sep 12, 2017).

mode pipeline break heat sink failure reactor facility pressure stop valve departure from nuclear boiling safety factor high-speed dump valves with discharge to atmosphere (HSDV-A)

Link for citing the article: Belozerov V.I., Zhuk M.M. Investigation of the small break conditions in the primary circuit of a VVER-1000 reactor. Izvestiya vuzov. Yadernaya Energetika. 2018, no. 3, pp. 137-147; DOI: https://doi.org/10.26583/npe.2018.3.12 (in Russian).

Open PDF file