Nodal models of Pressurized Water Reactor core for control purposes – A comparison study - Publikacja - MOST Wiedzy

Wyszukiwarka

Nodal models of Pressurized Water Reactor core for control purposes – A comparison study

Abstrakt

The paper focuses on the presentation and comparison of basic nodal and expanded multi-nodal models of the Pressurized Water Reactor (PWR) core, which includes neutron kinetics, heat transfer between fuel and coolant, and internal and external reactivity feedback processes. In the expanded multi-nodal model, the authors introduce a novel approach to the implementation of thermal power distribution phenomena into the multi-node model of reactor core. This implementation has the form of thermal power distribution coefficients which approximate the thermal power generation profile in the reactor. It is assumed in the model that the thermal power distribution is proportional to the axial distribution of neutron flux in the un-rodded and rodded reactor core regions, as a result of control rod bank movements. In the paper, the authors propose a methodology to calculate those power distribution coefficients, which bases on numerical solutions of the transformed diffusion equations for the un-rodded and rodded reactor regions, respectively. Introducing power distribution coefficients into the expanded multi-nodal model allows to achieve advanced capabilities that can be efficiently used in design and synthesis of more advanced and complex control algorithms for PWR reactor core, for instance in the field of reactor temperature distribution control.

Cytowania

  • 1 8

    CrossRef

  • 1 8

    Web of Science

  • 1 7

    Scopus

Cytuj jako

Pełna treść

pobierz publikację
pobrano 12 razy
Wersja publikacji
Accepted albo Published Version
Licencja
Creative Commons: CC-BY-NC-ND otwiera się w nowej karcie

Słowa kluczowe

Informacje szczegółowe

Kategoria:
Publikacja w czasopiśmie
Typ:
artykuł w czasopiśmie wyróżnionym w JCR
Opublikowano w:
NUCLEAR ENGINEERING AND DESIGN nr 322, strony 444 - 463,
ISSN: 0029-5493
Język:
angielski
Rok wydania:
2017
Opis bibliograficzny:
Puchalski B., Rutkowski T. A., Duzinkiewicz K.: Nodal models of Pressurized Water Reactor core for control purposes – A comparison study// NUCLEAR ENGINEERING AND DESIGN. -Vol. 322, (2017), s.444-463
DOI:
Cyfrowy identyfikator dokumentu elektronicznego (otwiera się w nowej karcie) 10.1016/j.nucengdes.2017.07.005
Bibliografia: test
  1. Dong, Z., Huang, X., Feng, J., Zhang, L., 2009. Dynamic model for control system design and simulation of a low temperature nuclear reactor. Nuclear Engineering and Design 239, 2141-2151. doi:10.1016/j.nucengdes.2009. 05.006. otwiera się w nowej karcie
  2. Dong, Z., Huang, X., Zhang, L., 2010. A nodal dynamic model for control system design and simulation of an MHTGR core. Nuclear Engineering and Design 240, 1251-1261. doi:10.1016/j.nucengdes.2009.12.032. otwiera się w nowej karcie
  3. Duderstadt, J.J., Hamilton, L.J., 1976. Nuclear reactor analysis. Wiley. otwiera się w nowej karcie
  4. Espinosa-Paredes, G., Polo-Labarrios, M.A., Espinosa-Martínez, E.G., Valle- Gallegos, E.d., 2011. Fractional neutron point kinetics equations for nuclear reactor dynamics. Annals of Nuclear Energy 38, 307-330. doi:10.1016/j. anucene.2010.10.012. otwiera się w nowej karcie
  5. Fazekas, C., Szederkényi, G., Hangos, K., 2007. A simple dynamic model of the primary circuit in VVER plants for controller design purposes. Nuclear Engineering and Design 237, 1071-1087. doi:10.1016/j.nucengdes.2006. 12.002. otwiera się w nowej karcie
  6. Fortum, VTT, 2017. Apros -Dynamic Process Simulation Software for Nuclear and Thermal Power Plant Applications. URL: http://www.apros.fi/en.
  7. Guimarães, L.N.F., Oliveira, N.d.S., Borges, E.M., 2008. Derivation of a nine variable model of a U-tube steam generator coupled with a three-element controller. Applied Mathematical Modelling 32, 1027-1043. doi:10.1016/j. apm.2007.02.022. otwiera się w nowej karcie
  8. Han, G.Y., 2000. A mathematical model for the thermal-hydraulic analysis of nuclear power plants. International Communications in Heat and Mass Transfer 27, 795-805. doi:10.1016/S0735-1933(00)00160-3. otwiera się w nowej karcie
  9. Kapernick, J.R., 2015. Dynamic Modeling of a Small Modular Reactor for Control and Monitoring. Ph.D. thesis. University of Tennessee.
  10. Karla, T., Tarnawski, J., Duzinkiewicz, K., 2015. Cross-platform real-time nu- clear reactor basic principle simulator, in: 2015 20th International Conference on Methods and Models in Automation and Robotics (MMAR), IEEE. pp. 1074-1079. doi:10.1109/MMAR.2015.7284028. otwiera się w nowej karcie
  11. Kerlin, T., 1978. Dynamic Analysis and Control of Pressurized Water Reactors, pp. 103-212. doi:10.1016/B978-0-12-012714-6.50008-8. otwiera się w nowej karcie
  12. Kulkowski, K., Kobylarz, A., Grochowski, M., Duzinkiewicz, K., 2015. Dynamic model of nuclear power plant steam turbine. Archives of Control Sciences 25, 65-86. doi:10.1515/acsc-2015-0005. otwiera się w nowej karcie
  13. Lewis, E.E., 2008. Fundamentals of Nuclear Reactor Physics. Elsevier. doi:10. 1016/B978-0-12-370631-7.X0001-0. otwiera się w nowej karcie
  14. Liu, X., 2015. Modeling and Simulation of a Prototypical Advanced Reactor. Ph.D. thesis. University of Tennessee.
  15. Naghedolfeizi, M., 1990. Dynamic Modeling of a Pressurized Water Reactor Plant for Diagnostics and Control. Ph.D. thesis. University of Tennessee. otwiera się w nowej karcie
  16. Nowak, T.K., Duzinkiewicz, K., Piotrowski, R., 2014a. Fractional neutron point kinetics equations for nuclear reactor dynamics Numerical solution investi- gations. Annals of Nuclear Energy 73, 317-329. doi:10.1016/j.anucene. 2014.07.001. otwiera się w nowej karcie
  17. Nowak, T.K., Duzinkiewicz, K., Piotrowski, R., 2014b. Numerical Solution of Fractional Neutron Point Kinetics Model in Nuclear Reactor. Archives of Control Sciences 24, 129-154. doi:10.2478/acsc-2014-0009. otwiera się w nowej karcie
  18. Nowak, T.K., Duzinkiewicz, K., Piotrowski, R., 2015. Numerical solution anal- ysis of fractional point kinetics and heat exchange in nuclear reactor. Nuclear Engineering and Design 281, 121-130. doi:10.1016/j.nucengdes.2014.11. 028. otwiera się w nowej karcie
  19. Perillo, S.R.P., 2010. Multi-Modular Integral Pressurized Water Reactor Control and Operational Reconfiguration for a Flow Control Loop. Ph.D. thesis. University of Tennessee. otwiera się w nowej karcie
  20. Puchalski, B., Duzinkiewicz, K., Rutkowski, T., 2015a. Multi-region fuzzy logic controller with local PID controllers for U-tube steam generator in nu- clear power plant. Archives of Control Sciences 25, 429-444. doi:10.1515/ acsc-2015-0028. otwiera się w nowej karcie
  21. Puchalski, B., Rutkowski, T., Tarnawski, J., Duzinkiewicz, K., 2015b. Compar- ison of tuning procedures based on evolutionary algorithm for multi-region fuzzy-logi PID controller for non-linear plant, in: 2015 20th International Conference on Methods and Models in Automation and Robotics (MMAR), IEEE. pp. 897-902. doi:10.1109/MMAR.2015.7283996. otwiera się w nowej karcie
  22. Puchalski, B., Rutkowski, T.A., Duzinkiewicz, K., 2016. Multi-nodal PWR reactor model -Methodology proposition for power distribution coefficients calculation, in: 2016 21st International Conference on Methods and Models in Automation and Robotics, MMAR 2016, IEEE. pp. 385-390. doi:10.1109/ MMAR.2016.7575166. otwiera się w nowej karcie
  23. Sharma, G., Bandyopadhyay, B., Tiwari, A., 2003. Spatial control of a large pressurized heavy water reactor by fast output sampling technique. IEEE Transactions on Nuclear Science 50, 1740-1751. doi:10.1109/TNS.2003. 818271. otwiera się w nowej karcie
  24. Sokolski, P., Rutkowski, T.A., Duzinkiewicz, K., 2016. Simplified, multiregional fuzzy model of a nuclear power plant steam turbine, in: 2016 21st Inter- national Conference on Methods and Models in Automation and Robotics (MMAR), IEEE. pp. 379-384. doi:10.1109/MMAR.2016.7575165. otwiera się w nowej karcie
  25. Tarnawski, J., Karla, T., 2016. Real-time simulation in non real-time envi- ronment, in: 2016 21st International Conference on Methods and Models in Automation and Robotics (MMAR), IEEE. pp. 577-582. doi:10.1109/MMAR. 2016.7575200. otwiera się w nowej karcie
  26. Tiwari, A., Banyopadhyay, B., Govindarajan, G., 1996. Spatial control of a large pressurized heavy water reactor. IEEE Transactions on Nuclear Science 43, 2440-2453. doi:10.1109/23.531794. otwiera się w nowej karcie
  27. Zhang, T., 2012. Comparison of Distributed Parameter and Multi-lump Models for a Pressurized Water Reactor Core. Ph.D. thesis. Arizona State University. otwiera się w nowej karcie
  28. Zhang, T., E. Holbert, K., 2013. Frequency Domain Comparison of Multi- lump and Distributed Parameter Models for Pressurized Water Reactor Cores. otwiera się w nowej karcie
  29. American Journal of Energy Research 1, 17-24. doi:10.12691/ajer-1-1-3. otwiera się w nowej karcie
Weryfikacja:
Politechnika Gdańska

wyświetlono 124 razy

Publikacje, które mogą cię zainteresować

Meta Tagi