Analysis of uncontrolled dilution of boric acid concentration in the reactor VVER-1000/320
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J. Hádek
Abstract
The paper gives a description of conservative analysis of initiating event associated with uncontrolled dilution of boric acid concentration in the reactor VVER-1000/320 of Temelín NPP. This event is included in the group of beyond design basis accidents. The aim of analysis is to determine also the time interval which is necessary for interventions leading to the deterrence of fuel damage. The failure of operator intervention to isolate dilution routes at intervals shorter than 30 min is assumed. Since the response of the whole NPP system influences the course of safety important parameters of the reactor core, the calculations were made by an externally coupled version of the 3D reactor dynamic code DYN3D and the thermohydraulic system code ATHLET. It is shown that, in addition to exceeding the DNBR limit of more than 99 min from the start of the transient, the remaining safety acceptance criteria will not be violated until the end of the calculation.
Kurzfassung
Das Papier gibt eine Beschreibung der konservativen Analyse des auslösenden Ereignisses im Zusammenhang mit der unkontrollierten Verdünnung der Borsäurekonzentration im Reaktor VVER-1000/320 des Kernkraftwerks Temelín. Dieses Ereignis ist Teil der Gruppe der Unfälle, die über die Auslegungskriterien hinausgehen. Ziel der Analyse ist es, auch das Zeitintervall zu bestimmen, das für Eingriffe zur Verhinderung von Brennstabschäden notwendig ist. Es wird davon ausgegangen, dass es dem Operator nicht gelingt, die Verdünnungswege innerhalb von 30 Minuten zu isolieren. Da die Reaktion des gesamten KKW-Systems den Verlauf der sicherheitstechnisch wichtigen Parameter des Reaktorkerns beeinflusst, wurden die Berechnungen mit einer extern gekoppelten Version des 3D-Reaktor-Dynamikcodes DYN3D und des thermohydraulischen Systemcodes ATHLET durchgeführt. Es wird gezeigt, dass außer der Überschreitung des DNBR-Grenzwertes von mehr als 99 Minuten ab Beginn der Transiente die verbleibenden Akzeptanzkriterien bis zum Ende der Berechnung nicht verletzt werden.
References
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© 2019, 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 2018
- Technical Contributions/Fachbeiträge
- Development of CASMO5 for VVER-1000/1200 analysis and preliminary validation using critical experiments
- C-PORCA 7: a nodal diffusion reactor calculation code to support off-line and on-line core analysis at Paks nuclear power plant
- Adaptation of the gas gap simplified model in DYN3D code to new types of fuel
- A procedure for verification of Studsvik's spent nuclear fuel code SNF
- Extension of nodal diffusion solver of Ants to hexagonal geometry
- VVER-1000 fuel assembly model in CAD-based unstructured mesh for MCNP6
- Fuel cycles with PK-3+ FAs for VVER-440 reactors
- Prospects for implementation of VVER nuclear fuel enriched above 5%
- Core loading optimisation in Slovak VVER-440 reactors
- Statistical evaluation of C-15 cycles in Paks NPP, based on measured in-core data
- Optimized 18-months low-leakage core loadings for uprated VVER-1000
- Leningrad NPP-2 start-up loss of power test and its simulation with use of KORSAR/GP code
- Assessment of the fuel assembly pin-by-pin model in the KORSAR/GP code
- Comparative thermohydraulic analyses of VVER 1000 active core for two different construction types of assemblies
- Analysis of uncontrolled dilution of boric acid concentration in the reactor VVER-1000/320
- Applied study on optimizing the final disposal of Loviisa NPP spent fuel assemblies
- Criticality safety analysis for GNS IQ® – The Integrated Quiver System for damaged fuel
- Neutron balance in two-component nuclear energy system
Articles in the same Issue
- Contents/Inhalt
- Contents
- Editorial
- Research on the reactor physics and reactor safety of VVER reactors – AER Symposium 2018
- Technical Contributions/Fachbeiträge
- Development of CASMO5 for VVER-1000/1200 analysis and preliminary validation using critical experiments
- C-PORCA 7: a nodal diffusion reactor calculation code to support off-line and on-line core analysis at Paks nuclear power plant
- Adaptation of the gas gap simplified model in DYN3D code to new types of fuel
- A procedure for verification of Studsvik's spent nuclear fuel code SNF
- Extension of nodal diffusion solver of Ants to hexagonal geometry
- VVER-1000 fuel assembly model in CAD-based unstructured mesh for MCNP6
- Fuel cycles with PK-3+ FAs for VVER-440 reactors
- Prospects for implementation of VVER nuclear fuel enriched above 5%
- Core loading optimisation in Slovak VVER-440 reactors
- Statistical evaluation of C-15 cycles in Paks NPP, based on measured in-core data
- Optimized 18-months low-leakage core loadings for uprated VVER-1000
- Leningrad NPP-2 start-up loss of power test and its simulation with use of KORSAR/GP code
- Assessment of the fuel assembly pin-by-pin model in the KORSAR/GP code
- Comparative thermohydraulic analyses of VVER 1000 active core for two different construction types of assemblies
- Analysis of uncontrolled dilution of boric acid concentration in the reactor VVER-1000/320
- Applied study on optimizing the final disposal of Loviisa NPP spent fuel assemblies
- Criticality safety analysis for GNS IQ® – The Integrated Quiver System for damaged fuel
- Neutron balance in two-component nuclear energy system
