Power transient calculations with VERONA
-
M. Horváth
Abstract
In Paks Nuclear Power Plant the VERONA core monitoring system is being used. In this article there is a new way presented to determine the precision of VERONA calculations during power transients. In that case VERONA uses a sophisticated method to follow the changes of the core. During a transient many online measured parameters cannot be used to determine the precision of the model due to their insufficient accuracy. However the change of the calculated boric acid concentration can be used for this purpose. If there is no inlet of boric acid or clean condense during the transient, at the end, the calculated boric acid concentration value should be equal to that, which was determined at the beginning of the power transient. However there is often a difference between these two values, which difference denotes the precision of the calculating model. This precision is heavily influenced by the heat conduction model, used between the fuel and the coolant. During the work five different conduction models were investigated, to determine which one offers more precise results. There was also a way examined to determine the uncertainty of the fuel temperature coefficient based on the outcome of the examined power transitions.
Kurzfassung
Im Kernkraftwerk Paks wird das Kernüberwachungssystem VERONA eingesetzt. In diesem Artikel wird ein neuer Weg vorgestellt, um die Genauigkeit von VERONA-Berechnungen bei Leistungstransienten zu bestimmen. Dazu verwendet VERONA eine ausgeklügelte Methode, um die Veränderungen des Kerns zu verfolgen. Während einer Transiente können viele online gemessene Parameter aufgrund ihrer ungenügenden Genauigkeit nicht zur Bestimmung der Genauigkeit des Modells herangezogen werden. Dazu kann jedoch die Änderung der berechneten Borsäurekonzentration genutzt werden. Wenn während der Transiente kein Borsäureeinlass oder sauberes Kondensat vorhanden ist, sollte der errechnete Borsäurekonzentrationswert am Ende gleich demjenigen sein, der zu Beginn der Leistungstransiente ermittelt wurde. Allerdings gibt es oft einen Unterschied zwischen diesen beiden Werten, der die Genauigkeit des Berechnungsmodells angibt. Diese Präzision wird stark durch das Wärmeleitungsmodell beeinflusst, das zwischen dem Brennstoff und dem Kühlmittel eingesetzt wird. Während der Arbeit wurden fünf verschiedene Leitungsmodelle untersucht, um festzustellen, welches genauere Ergebnisse liefert. Es wurde auch eine Möglichkeit untersucht, die Unsicherheit des Brennstofftemperaturkoeffizienten anhand der Ergebnisse der untersuchten Leistungsübergänge zu bestimmen.
References
1 Pos, I.; Kalya, Z.; Parko, T.; Patai Szabo, S.: New models in VERONA 7.0 system. 25. AER symposium 2015; Balatongyoeroek (Hungary) 13–16 Oct 2015 and Kerntechnik47 (2016) 380–386Search in Google Scholar
2 Henry, A. F.: The application of reactor kinetics to the analysis of experiments. Nuclear Science and Engineering3 (1958) 52–7010.13182/NSE58-1Search in Google Scholar
3 Kulacsy, K.; Griger, A.; Keresztury, A.; Molnar, A.; Panka, I.; Slonszki, E.; Somfai, B.: Extensive validation of the code FUROM based on IFPE database. International conference on VVER fuel performance, modeling and experimental support; Helena Resort (Bulgaria); 28. Sep- 2 Oct 2009Search in Google Scholar
4 Fedotov, P.; Kumarchev, A.; Nechaeva, O.; Novikov, V.; Salatov, A.; Sypchenko, M.: VVERMIR – LOCA/72 Experiment post-test simulation with RAPTA 5.2 code. 11. International conference on WWER fuel performance, modeling and experimental support Proceedings, Varna (Bulgaria), 26 Sep – 3 Oct 2015Search in Google Scholar
© 2018, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Editorial
- Research on the reactor physics and reactor safety of VVER reactors – AER Symposium 2017
- Technical Contributions/Fachbeiträge
- SIMULATE5-HEX extension for VVER analyses
- Application of discontinuity factors and group constants generated by SERPENT in the KIKO3 DMG code
- “Full-Core” VVER-440 extended calculation benchmark
- Calculation of “full core” VVER-1000 benchmark
- Study of neutron-physical characteristics of VVER-1200 considering feedbacks using MCU Monte Carlo code
- Advantages of VVER-440 fuel cycles with new fuel assemblies
- A neutronics feasibility study on utilization of a thinned cladding fuel design at Loviisa NPP
- Investigation of fuel cycles containing Generation IV reactors and VVER-1200 reactors
- Calculations of spent fuel isotopic composition for fuel rod from VVER-440 fuel assembly benchmark using several evaluated nuclear data libraries
- Simulation of standard temperature control indications at the outlet of a fuel assembly of VVER1000 reactor of Rostov NPP unit No. 2
- Power transient calculations with VERONA
- Physical startup tests calculations for Dukovany NPP using MOBY-DICK macrocode
- Renewing the refueling neutron monitoring and reactivity measurement systems at Paks NPP
- Hot channel calculation methodologies in case of VVER-1000/1200 reactors
- Contribution to the validation of the VVER-1000 Temelin NPP computing model for the ATHLET/DYN3D coupled codes
- Simulation of a hypothetical MSLB core transient in VVER-1000 with several stuck rods
Articles in the same Issue
- Contents/Inhalt
- Contents
- Editorial
- Research on the reactor physics and reactor safety of VVER reactors – AER Symposium 2017
- Technical Contributions/Fachbeiträge
- SIMULATE5-HEX extension for VVER analyses
- Application of discontinuity factors and group constants generated by SERPENT in the KIKO3 DMG code
- “Full-Core” VVER-440 extended calculation benchmark
- Calculation of “full core” VVER-1000 benchmark
- Study of neutron-physical characteristics of VVER-1200 considering feedbacks using MCU Monte Carlo code
- Advantages of VVER-440 fuel cycles with new fuel assemblies
- A neutronics feasibility study on utilization of a thinned cladding fuel design at Loviisa NPP
- Investigation of fuel cycles containing Generation IV reactors and VVER-1200 reactors
- Calculations of spent fuel isotopic composition for fuel rod from VVER-440 fuel assembly benchmark using several evaluated nuclear data libraries
- Simulation of standard temperature control indications at the outlet of a fuel assembly of VVER1000 reactor of Rostov NPP unit No. 2
- Power transient calculations with VERONA
- Physical startup tests calculations for Dukovany NPP using MOBY-DICK macrocode
- Renewing the refueling neutron monitoring and reactivity measurement systems at Paks NPP
- Hot channel calculation methodologies in case of VVER-1000/1200 reactors
- Contribution to the validation of the VVER-1000 Temelin NPP computing model for the ATHLET/DYN3D coupled codes
- Simulation of a hypothetical MSLB core transient in VVER-1000 with several stuck rods
