The effect of boron dilution transient on the VVER-1000 reactor core using MCNP and COBRA-EN codes
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N. Jafari
Abstract
In this paper, the effect of boron dilution transient, as a consequence of the malfunction of the boron control system, was investigated in a VVER-1000 reactor, and then an appropriate setpoint was determined for the actuation of the emergency protection system to the reactor shutdown. In order to simulate the boron dilution, first, the whole reactor core was simulated by MCNPX code to compute the radial and axial power distribution. Then, the COBRA-EN code was employed using calculated power distribution for analyzing the thermal-hydraulic of hot fuel assembly and for extracting the safety parameters. For the safe operation of the reactor, certain parameters must be in defined specified ranges. Comparison between our results and FSARs data shows that the present modeling provides a good prediction of boron dilution transient with the maximum relative difference about 4%.
Kurzfassung
In diesem Beitrag wurde der Effekt einer Borverdünnungstransiente infolge einer Funktionsstörung des Borkontrollsystems in einem WWER-1000-Reaktor untersucht. Dann wurde ein geeigneter Sollwert für den Start des Notfallschutzsystems zur Reaktorabschaltung bestimmt. Zur Simulation der Borverdünnung wurde zuerst der vollständige Reaktorkern mit Hilfe des MCNPX-Codes simuliert zur Berechnung der radialen und axialen Leistungsverteilung. Dann wurde der COBRA-EN-Code mit Hilfe berechneter Leistungsverteilung zur Analyse der Thermohydraulik der Brennelemente und zur Gewinnung der Sicherheitsparameter angewandt. Für den sicheren Betrieb des Reaktors müssen sich bestimmte Parameter in definierten Bereichen befinden. Der Vergleich zwischen unseren Ergebnissen und den Daten des endgültigen Sicherheitsberichts zeigt, dass die derzeitige Modellierung eine gute Vorhersage einer Borverdünnungstransiente liefert mit der maximalen relativen Abweichung von etwa 4%.
References
1 Atomic Energy Organization of Iran, 2003. Album of Neutron and Physical Characteristics of the 1st Loading of Boushehr Nucl. Plant, Technical Report, Tehran, IranSearch in Google Scholar
2 Final Safety Analysis Report of Bushehr's VVER-1000 Reactor. Ministry of Russian Federation of Atomic Energy, Moscow, 2003, Chapter 4 and 15Search in Google Scholar
3 Talebi, S.; Kazeminejad, H.: A mathematical approach to predict dryout in a rod bundle. Nuclear Engineering and Design249 (2012) 348–35610.1016/j.nucengdes.2012.04.016Search in Google Scholar
4 AP1000 Design Control Document, NRC, 2008, Revision 14, Chapter 15Search in Google Scholar
5 Talebi, S.; Kazeminejad, H.; Davilua, H.: Prediction of dryout and post dryout wall temperatures using film thickness model. Nuclear Engineering and Design244 (2012) 73–8210.1016/j.nucengdes.2011.12.032Search in Google Scholar
6 Arias, F. J.: Boron dilution effect on boiling heat transfer with special reference to nuclear reactors technology. Annals of Nuclear Energy36 (2009) 1382–138510.1016/j.anucene.2009.06.018Search in Google Scholar
7 Rohde, U.; Elkin, I.; Kalinenko, V.: Analysis of a boron dilution accident for WWER-440 combining the use of the codes DYN3D and SiTap. Nuclear Engineering and Design170 (1997) 95–9910.1016/S0029-5493(97)00016-2Search in Google Scholar
8 Jimenez, G.; Herrero, J. J.; Gommlich, A.; Kliem, S.; Cuervo, D.; Jimenez, J.: Boron dilution transient simulation analyses in a PWR with neutronics/thermal-hydraulics coupled codes in the NURISP project Design. Annals of Nuclear Energy84 (2014) 86–9710.1016/j.anucene.2014.11.002Search in Google Scholar
9 Ivanov, K. N.; Grundmann, U.; Mittag, S.; Rohde, U.: Comparative study of a boron dilution scenario in VVER reactors. Annals of Nuclear Energy26 (1999) 1331–133910.1016/S0306-4549(99)00018-3Search in Google Scholar
10 Pandey, Y. K.; Chauhan, A.: Fuel Management of VVER-1000 Reactors of Kudankulam Nuclear Power Plant. Nuclear Power of Corporation of India Limited, Nabhikiya Urja Bhavan, Anushaktinagar, Mumbai (2009) 400094Search in Google Scholar
11 Hainoun, A.; Haj Hassan, H.; Ghazi, N.: Determination of major kinetic parameters of the Syrian MNSR for different fuel loading using Monte Carlo technique. Annals of Nuclear Energy36 (2009) 1663–166710.1016/j.anucene.2009.09.010Search in Google Scholar
© 2017, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- SIMULATE-3 K coupled code applications
- Application of the new IAPWS Guideline on the fast and accurate calculation of steam and water properties with the Spline-Based Table Look-Up Method (SBTL) in RELAP-7
- Simulation of water hammer phenomena using the system code ATHLET
- New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses
- Sensitivity analysis for thermo-hydraulics model of a Westinghouse type PWR: verification of the simulation results
- Calculation of the fuel temperature coefficient of reactivity considering non-uniform radial temperature distribution in the fuel rod
- The effect of boron dilution transient on the VVER-1000 reactor core using MCNP and COBRA-EN codes
- Modelling of the spent fuel heat-up in the spent fuel pools using one-dimensional system codes and CFD codes
- Optimization and analysis of the effects of physical parameters in a TRIGA-ADSR
- A comparison study for mass attenuation coefficients of some amino acids using MCNP code
- Validation of radioactive isotope activity measurement in homogeneous waste drum using Monte Carlo codes
- Study of the response reduction of LiF:Mg, Ti dosimeter for high dose dosimetry
- Non-contact micro mass evaluation method using an X-ray microscope
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- SIMULATE-3 K coupled code applications
- Application of the new IAPWS Guideline on the fast and accurate calculation of steam and water properties with the Spline-Based Table Look-Up Method (SBTL) in RELAP-7
- Simulation of water hammer phenomena using the system code ATHLET
- New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses
- Sensitivity analysis for thermo-hydraulics model of a Westinghouse type PWR: verification of the simulation results
- Calculation of the fuel temperature coefficient of reactivity considering non-uniform radial temperature distribution in the fuel rod
- The effect of boron dilution transient on the VVER-1000 reactor core using MCNP and COBRA-EN codes
- Modelling of the spent fuel heat-up in the spent fuel pools using one-dimensional system codes and CFD codes
- Optimization and analysis of the effects of physical parameters in a TRIGA-ADSR
- A comparison study for mass attenuation coefficients of some amino acids using MCNP code
- Validation of radioactive isotope activity measurement in homogeneous waste drum using Monte Carlo codes
- Study of the response reduction of LiF:Mg, Ti dosimeter for high dose dosimetry
- Non-contact micro mass evaluation method using an X-ray microscope
