Startseite Assessment of the fuel assembly pin-by-pin model in the KORSAR/GP code
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Assessment of the fuel assembly pin-by-pin model in the KORSAR/GP code

  • A. I. Sinegribova , M. A. Uvakin und M. A. Bykov
Veröffentlicht/Copyright: 27. August 2019
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Kerntechnik
Aus der Zeitschrift Kerntechnik Band 84 Heft 4

Abstract

The purpose of this paper is the assessment of the FA pin-by-pin model available in the KORSAR/GP code. The paper presents the results of the simulation of an “Control rod ejection” accident. This accident is of interest due to the power redistribution in the FAs effected by the ejection. The safety margins are calculated with the use of the “hot channel” model for NPP safety analyses. Only the fuel rods of the core with the highest power are taking into account. The absence of the change of the pin power distribution for the FA during the transient is compensated by increased conservatism. In this paper, the maximum fuel rod power was obtained using the pin-by-pin model.

Kurzfassung

Dieser Beitrag präsentiert eine Bewertung des Brennelement-Pin-by-Pin-Modells des Programms KORSAR/GP anhand einer Berechnung des Szenarios Steuerstabauswurf. Dieses Szenario ist aufgrund der durch den Auswurf verursachten Energieumverteilung in den Brennelementen von Interesse. Die Sicherheitsmargen werden mit Hilfe des „Heißkanal“-Modells für die Sicherheitsanalysen von Kernkraftwerken berechnet. Es werden nur die Brennstäbe des Kerns mit der höchsten Leistung berücksichtigt. Das Fehlen einer Änderung der Pin-Leistungsverteilung für das Brennelement während der Transienten wird durch erhöhten Konservatismus kompensiert. In diesem Beitrag wurde die maximale Brennstableistung mit dem Pin-by-Pin-Modell ermittelt.

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References

1 Gordienko, P. V.: Transient neutron-physics simulation using pin-by-pin nodalization for VVER NPP. Dissertation abstract for the candidate of engineering sciences degree Moscow, 2014Suche in Google Scholar

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4 Panka, I.; Hegyi, Gy.; Kereszturi, A.; Maraczy, Cs.; Temesvari, E.: Hot channel calculation methodologies in case of VVER-1000/1200 reactors. Kerntechnik83 (2018) 365–37510.3139/124.110899Suche in Google Scholar

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7 Sinegribova, A. I.; Petkevich, I. G.: Development of the fuel assembly pin-by-pin model in the code KORSAR/GP. Proceedings of the 10th International Conference “Safety Assurance of NPP with WWER. Russia, Moscow region, Podolsk, 2017Suche in Google Scholar

8 Sinegribova, A. I.; Uvakin, M. A.: Application of fuel assembly pin-by-pin model developed by code KORSAR/GP for VVER NPP transient simulation. Problems of Atomic Science and Technology. Series: Nuclear and Reactor Constants, 4 (2017) (in Russian)Suche in Google Scholar

Received: 2019-02-15
Published Online: 2019-08-27
Published in Print: 2019-09-16

© 2019, Carl Hanser Verlag, München

Artikel in diesem Heft

  1. Contents/Inhalt
  2. Contents
  3. Editorial
  4. Research on the reactor physics and reactor safety of VVER reactors – AER Symposium 2018
  5. Technical Contributions/Fachbeiträge
  6. Development of CASMO5 for VVER-1000/1200 analysis and preliminary validation using critical experiments
  7. C-PORCA 7: a nodal diffusion reactor calculation code to support off-line and on-line core analysis at Paks nuclear power plant
  8. Adaptation of the gas gap simplified model in DYN3D code to new types of fuel
  9. A procedure for verification of Studsvik's spent nuclear fuel code SNF
  10. Extension of nodal diffusion solver of Ants to hexagonal geometry
  11. VVER-1000 fuel assembly model in CAD-based unstructured mesh for MCNP6
  12. Fuel cycles with PK-3+ FAs for VVER-440 reactors
  13. Prospects for implementation of VVER nuclear fuel enriched above 5%
  14. Core loading optimisation in Slovak VVER-440 reactors
  15. Statistical evaluation of C-15 cycles in Paks NPP, based on measured in-core data
  16. Optimized 18-months low-leakage core loadings for uprated VVER-1000
  17. Leningrad NPP-2 start-up loss of power test and its simulation with use of KORSAR/GP code
  18. Assessment of the fuel assembly pin-by-pin model in the KORSAR/GP code
  19. Comparative thermohydraulic analyses of VVER 1000 active core for two different construction types of assemblies
  20. Analysis of uncontrolled dilution of boric acid concentration in the reactor VVER-1000/320
  21. Applied study on optimizing the final disposal of Loviisa NPP spent fuel assemblies
  22. Criticality safety analysis for GNS IQ® – The Integrated Quiver System for damaged fuel
  23. Neutron balance in two-component nuclear energy system
https://doi.org/10.3139/124.190027

Artikel in diesem Heft

  1. Contents/Inhalt
  2. Contents
  3. Editorial
  4. Research on the reactor physics and reactor safety of VVER reactors – AER Symposium 2018
  5. Technical Contributions/Fachbeiträge
  6. Development of CASMO5 for VVER-1000/1200 analysis and preliminary validation using critical experiments
  7. C-PORCA 7: a nodal diffusion reactor calculation code to support off-line and on-line core analysis at Paks nuclear power plant
  8. Adaptation of the gas gap simplified model in DYN3D code to new types of fuel
  9. A procedure for verification of Studsvik's spent nuclear fuel code SNF
  10. Extension of nodal diffusion solver of Ants to hexagonal geometry
  11. VVER-1000 fuel assembly model in CAD-based unstructured mesh for MCNP6
  12. Fuel cycles with PK-3+ FAs for VVER-440 reactors
  13. Prospects for implementation of VVER nuclear fuel enriched above 5%
  14. Core loading optimisation in Slovak VVER-440 reactors
  15. Statistical evaluation of C-15 cycles in Paks NPP, based on measured in-core data
  16. Optimized 18-months low-leakage core loadings for uprated VVER-1000
  17. Leningrad NPP-2 start-up loss of power test and its simulation with use of KORSAR/GP code
  18. Assessment of the fuel assembly pin-by-pin model in the KORSAR/GP code
  19. Comparative thermohydraulic analyses of VVER 1000 active core for two different construction types of assemblies
  20. Analysis of uncontrolled dilution of boric acid concentration in the reactor VVER-1000/320
  21. Applied study on optimizing the final disposal of Loviisa NPP spent fuel assemblies
  22. Criticality safety analysis for GNS IQ® – The Integrated Quiver System for damaged fuel
  23. Neutron balance in two-component nuclear energy system
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