Izvestiya vuzov. Yadernaya Energetika

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

Heat exchange control in the VVER core using Petri nets

3/19/2020 2020 - #01 Modelling processes at nuclear facilities

Kachur S.A.

DOI: https://doi.org/10.26583/npe.2020.1.13

UDC: 621.039.56

The aim of the study is to develop a method for identifying the boiling process in the channels of the VVER core and a model of the information-measuring system to be used on the basis of Petri nets for more efficient heat exchange control in the reactor core during emergencies.

In this case, two tasks are solved: (1) the creation of a method for identifying the «hottest» steam distribution spots in the mixing chamber; and (2) the development of a model for controlling the information-measuring system in the process of this identification, as a component of the reactor power control system.

It is proposed to identify the boiling process in the reactor core according to the information of the optical information-measuring system located along the perimeter of the mixing chamber. The developed method for controlling the steam void in the mixing chamber is called the method of the «hottest» spots for a nuclear reactor. This method is based on solving systems of linear equations that describe various schemes for using information from optical sensors.

A scheme of an adaptive heat transfer control system of a nuclear reactor and a model based on the method of the «hottest» spots and expansion of Petri nets have been developed, taking into account the stochastic nature of the heat exchange process, which makes it possible to increase by an order of magnitude the rate of response and decision-making during emergencies.

References

  1. Kachur S.A. Control of Nuclear Reactor Power Distribution Based on Petri Nets. Energeticheskie Ustanovki i Tekhnologii, 2019, v. 5, no. 1, pp. 14-20 (in Russian).
  2. Kirillov P.L. Reference Book upon Heat0Hydraulic Accounts in Nuclear Power Industry. Moscow. IzdAT Publ., 2010, 771 p. (in Russian).
  3. Sharoevskij I.G., Domashev E.D., Arhipova A.P. Verification Method for Beginning Boiling up of Coolant in the Channels of Nuclear Reactor. Promyshlennaya Teplotehnira, 2001, v. 23, no. 4-5, pp. 114-121 (in Russian).
  4. Sharoevskij I.G. Recognition of the Modes of Flow of Diphasic Stream in the Channels of Nuclear Reactor on Noises of Technological Parameters. Promyshlennaya Teplotehnira, 2000, v. 22, no 1, pp. 53-59 (in Russian).
  5. Kovehkaya M.M. Crisis of Heat Exchange in the Bunches of Bars with Twirling of Stream. Promyshlennaya Teplotehnira, 2009, v. 31, no. 5, pp. 50-55 (in Russian).
  6. Kachur S.A. Diagnostics of the Crisis State of VVER Reactor Based on the Channel Steaming Model. Izvestiya vuzov. Yadernaya Energetika, 2019, no. 1, pp. 41-50. DOI: https://doi.org/ 10.26583/npe.2019.1.04 (in Russian).
  7. Kachur S.A. Prediction of Boiling Modes in a Steam Generating Channel Based on Spectral Analysis of Acoustic Noise. Energeticheskie Ustanovki i Tekhnologii, 2018, v. 4, no. 3, pp. 24-29 (in Russian).
  8. Gusev S.S. Safe control nuclear power plants. Elektrooborudovanie: Ekspluataciya i Remont, 2017, no. 12, pp. 42-45 (in Russian).
  9. Kachur S.А. Model of Stochastic Systems and Their Combination on the Basis of Petry Nets. Problemy Upravleniya i Informatiki, 2002, no.1, pp. 93-98 (in Russian).
  10. Korneeva T.V. Fiber-Optic Sensors: development, application. Pribory i Sistemy. Upravlenie, Kontrol’, Diagnostika , 2016, no. 7, pp. 25-41 (in Russian).
  11. Vesnin V.L., Chetorijskij A.A., Ekke V. Test and Measurement Systems Based on Fiber-Optic Bragg’ Sensors. Radiotehnika i Electronika, 2005, v. 50, no. 6, pp. 751-758 (in Russian).
  12. Suslov V.A. Flow Modes of Two-Phase Fluidized Flow in Pipes. Teplovye Processy v Tekhnike, 2012, v. 4, no. 12, pp. 539-543 (in Russian).
  13. Dolinskij A.A., Sharaevskij I.G., Fialko N.M. Methodology for Recognition and Verification of Crisis of Heat Transfer in Rod Assemblies. Promyshlennaya Teplotekhnika, 2005, v. 27, no. 6, pp. 66-80 (in Russian).
  14. Popov I.A., Domashev E.D., Sychev E.N., Zhuravlev A.A. Experimental Setup and Automated System of Data Collection and Processing for Simulation of Emergency Thermal-Hydraulic Processes. Promyshlennaya Teplotekhnika, 2007, v. 29, no. 2, pp. 62-68 (in Russian).
  15. Leontiev A.I., Olimpiev V.V. The Influence of Intensification of Heat Transfer on the Thermohydraulic Properties of Channels. Teplofizika Vysokikh Temperatur, 2007, v. 45, no. 6, pp. 925-953. DOI: https://doi.org/10.1134/S0018151X07060168 (in Russian).
  16. Kirillov P.L. New Methods of Intensification of Surface Heat Exchange with Boiling Water. Atomnaya Tekhnika za Rubezhom, 2005, no. 10, pp. 3-7 (in Russian).
  17. Kachur S.А., Bogma A.C. Modification of Automatic Control Systems Based on Statistical and Neural Network Methods. Energeticheskie Ustanovki i Tekhnologii, 2018, v. 4, no. 1, pp. 50-55 (in Russian).

nuclear reactor heat exchange steam void identification control Petri nets