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Statistical evaluation of C-15 cycles in Paks NPP, based on measured in-core data

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Published/Copyright: August 27, 2019
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Kerntechnik
From the journal Kerntechnik Volume 84 Issue 4

Abstract

After a preparation period with VERONA upgrade and lead test assembly program, a new fuel type was introduced at MVM Paks NPP Ltd. This 4.7% average uranium enriched assembly type, together with the former 4.2% uranium enriched fuels, allowed us to lengthen the operating cycles to 15 months. Both fuel types contain gadolinium burnable poison, in six and three pins respectively. All of the four units have been converted to the C15 cycles, and have been being operated without any problems in the last few years. In this paper the test results of core design code HELIOS/C-PORCA, which is the basic model of VERONA, are outlined. C15 cycles were entirely investigated with the comprehensive study of measured and predicted (calculated) values of different reactor states. At first step, in order to prove the capability of the reactivity calculation of the nodal diffusion model, critical boric acid concentrations of different burnup and start up states were calculated and compared with measured values. During the next step of the verification process local in-core parameters were investigated. Measured neutron flux distributions (SPND signals) and coolant outlet temperatures were examined. SPND and thermo couple (TC) signals were predicted as a part of the monitoring system. The results of statistical investigations (average differences and standard deviations) for the applied fuel types are also presented.

Kurzfassung

Nach einer Vorbereitungsphase wurde von MVM Paks NPP Ltd. neue Brennelemente mit Uran mit einer durchschnittlichen Anreicherung von 4,7% im Brennstoff in den bestehenden Kern mit Brennelementen mit der bisherigen 4,2% Anreicherung eingesetzt. Damit war eine Verlängerung des Betriebszyklus auf 15 Monate möglich. Beide Brennelementtypen enthalten in sechs bzw. drei Stäben das abbrennbare Neutronengift Gadolinium. Alle vier Kraftwerke am Standort Paks wurden auf diese C15-Zyklen umgestellt und sind in den letzten Jahren problemlos betrieben worden. In diesem Beitrag werden die Ergebnisse der Anwendung des Core-Design-Codes HELIOS/C-PORCA, der das Basismodell des Programms VERONA darstellt, beschrieben. Die C15-Zyklen wurden dabei vollständig, d.h. mit einem Vergleich von gemessenen und berechneten Werten verschiedener Reaktorzustände, untersucht. Um die Leistungsfähigkeit der Reaktivitätsberechnung des nodalen Diffusionsmodells nachzuweisen, wurden im ersten Schritt kritische Borsäurekonzentrationen verschiedener Brenn- und Anfahrzustände berechnet und mit Messwerten verglichen. Im nächsten Schritt des Verifikationsprozesses wurden lokale In-Core-Parameter untersucht. Gemessene Neutronenflussverteilungen (SPND-Signale) und Kühlmittelaustrittstemperaturen wurden untersucht. SPND- und Thermoelement (TC)-Signale wurden als Teil des Überwachungssystems berechnet. Die Ergebnisse statistischer Untersuchungen (durchschnittliche Unterschiede und Standardabweichungen) für die verwendeten Brennelementtypen werden ebenfalls dargestellt.

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References

1 Simeonov, T.: HELIOS v1.12 Release Notes. 2012Search in Google Scholar

2 Pós, I.; Kálya, Z., Patai-Szabó, S., Parkó, T.: New Models in VERONA 7.0 System. 25th Symposium of AER, Balatongyörök, Hungary, October 2015 10.3139/124.110704Search in Google Scholar

3 Végh, J.; Pós, I.; Horváth, Cs.; Kálya, Z.; Parkó, T.; Ignits, M.: VERONA V6.22 – An anhanced reactor analysis tool applied tool for continous core parameter monitoring at Paks NPP. Nuclear Engineering and Design292 (2015) 261–27610.1016/j.nucengdes.2015.06.006Search in Google Scholar

4 Parkó, T.; Pós, I.; Horváth, M.: Test results of new VERONA models based on measured in-core data. 25th Symposium of AER, Balatongyörök, Hungary, October 2015Search in Google Scholar

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Received: 2019-01-15
Published Online: 2019-08-27
Published in Print: 2019-09-16

© 2019, Carl Hanser Verlag, München

Articles in the same Issue

  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/10.3139/124.190028

Articles in the same Issue

  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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