Analyses of Beyond Design Basis Accident Homogeneous Boron Dilution Scenarios
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A. Keresztúri
Abstract
Homogeneous boron dilution scenarios in a VVER-440 reactor were analyzed using the coupled KIKO3D-ATHLET code. The scenarios are named “homogeneous” because of the very slow dilution caused by a rupture in the heat exchanger of the makeup system. In the case of the presented Beyond Design Basis Accident (BDBA) scenarios, it was supposed that after 30 min the operator doesn't stop the dilution, violating the prescriptions of the Limits and Conditions document. Without the presented analyses, a significant contribution of the homogeneous boron dilution to the Core Damage Frequency (CDF) had to be assumed in the Probabilistic Safety Analyses (PSA). According to the combined results of the presented deterministic and probabilistic analyses, the final conclusion is that boron dilution transients don't give significant contribution to the CDF for the investigated VVER-440 NPP.
Kurzfassung
In probabilistischen Sicherheitsanalysen für WWER-440 Reaktoren werden für auslegungsüberschreitende Störfallszenarien, wie z. B. Lecks im Wärmetausche des Makeupsystems, Einflüsse der homogenen Borverdünnung auf die Kernschadenshäufigkeit berücksichtigt, sofern keine deterministischen Untersuchungen vorliegen. Mit den in diesem Beitrag vorgestellten deterministischen und probabilistischen Berechnungen dieser Szenarien, u. a. mit dem gekoppelten Programmsystem KIKO3D-ATHLET, konnte gezeigt werden, dass diese Störfalltransienten keinen signifikanten Beitrag zur Kernschadenshäufigkeit in der hier untersuchten Anlage liefern.
References
1 Keresztúri, A.; Hegyi, Gy.; Maráczy, Cs.; Panka, I.; Telbisz, M.; Trosztel, I.; Hegedűs, Cs.: Development and validation of the three dimensional dynamic code-KIKO3D. Annals of Nuclear Energy30 (2003) 93–12010.1016/S0306-4549(02)00043-9Suche in Google Scholar
2 Keresztúri, A.; Hegyi, Gy.; Maráczy, Cs.: Experience in simulation of startup measurements in the VVER-440 reactor using the KIKO3D code. Joint International Topical Meeting on Mathematics & Computation and Supercomputing in Nuclear Applications (M&C+ SNA 2007), Monterey, California, April 15–19, 2007, CD-ROM, ANS, LaGrange Park, IL (2007)Suche in Google Scholar
3 Keresztúri, A.; Hegyi, Gy.; Korpás, L.; Maráczy, Cs.; Makai, M.; Telbisz, M.: General features and validation of the recent KARATE-440 code system. International Journal of Nuclear Energy Science and Technology5 (2010) 207–23810.1504/IJNEST.2010.033476Suche in Google Scholar
4 Lerchl, G.; Austregesilo, H.: ATHLET Mod2 Cycle A, User's Manual. GRS (2003)Suche in Google Scholar
5 Langenbuch, S.; Velkov, K.; Kliem, S.; Rhode, U.; Lizorkin, M.; Hegyi, Gy.; Keresztúri, A.: Development of coupled systems of 3D neutronics and fluid-dynamic system codes and their application for safety analysis. Eurosafe 2000, Cologne, Germany, 6–7 Nov. 2000Suche in Google Scholar
6 Hegyi, Gy.; Keresztúri, A.; Trosztel, I.: VVER-440 Steam Line Break Analysis by Means of the Coupled KIKO3D/ATHLET Code System. The 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13), Kanazawa City, Ishikawa Prefecture, Japan, September 27–October 2, 2009 paper N13P1244Suche in Google Scholar
7 Hegyi, Gy.; Hordósy, G.; Keresztúri, A.; Maráczy, Cs.; Panka, I.; Telbisz, M.; Trosztel, I.: Development of the Methodology of the Safety Analysis Performed by the Coupled KIKO3D/ATHLET Code System in VVER-440 Type NPP. Paper No. ICONE12-49300 (2004)10.1115/ICONE12-49300Suche in Google Scholar
8 Standard Review Plan for the Review of Safety Analysis Report of Nuclear Power Plants, US NRC, 1979Suche in Google Scholar
© 2015, Carl Hanser Verlag, München
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Editorial
- Research on the reactor physics and reactor safety of VVER reactors – AER Symposium 2014
- Technical Contributions/Fachbeiträge
- Assessment of the uncertainties of MULTICELL calculations by the OECD NEA UAM PWR pin cell burnup benchmark
- Development of codes and KASKAD complex
- Applying full multigroup cell characteristics from MCU code to finite difference calculations of neutron field in VVER core
- Calculations of 3D full-scale VVER fuel assembly and core models using MCU and BIPR-7A codes
- An analysis of reactivity prediction during the reactor start-up process
- Experimental and computational investigations of heat and mass transfer of intensifier grids
- Implementation of CFD module in the KORSAR thermal-hydraulic system code
- Numerical and experimental investigation of 3D coolant temperature distribution in the hot legs of primary circuit of reactor plant with WWER-1000
- Analyses of Beyond Design Basis Accident Homogeneous Boron Dilution Scenarios
- Analysis of heterogeneous boron dilution transients during outages with APROS 3D nodal core model
- Prospects of subcritical molten salt reactor for minor actinides incineration in closed fuel cycle
- Usage of burnt fuel isotopic compositions from engineering codes in Monte-Carlo code calculations
- Neutron-kinetic and thermo-hydraulic uncertainties in the study of Kalinin-3 benchmark
- Inter-assembly gap deviations in VVER-1000: Accounting for effects on engineering margin factors
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Editorial
- Research on the reactor physics and reactor safety of VVER reactors – AER Symposium 2014
- Technical Contributions/Fachbeiträge
- Assessment of the uncertainties of MULTICELL calculations by the OECD NEA UAM PWR pin cell burnup benchmark
- Development of codes and KASKAD complex
- Applying full multigroup cell characteristics from MCU code to finite difference calculations of neutron field in VVER core
- Calculations of 3D full-scale VVER fuel assembly and core models using MCU and BIPR-7A codes
- An analysis of reactivity prediction during the reactor start-up process
- Experimental and computational investigations of heat and mass transfer of intensifier grids
- Implementation of CFD module in the KORSAR thermal-hydraulic system code
- Numerical and experimental investigation of 3D coolant temperature distribution in the hot legs of primary circuit of reactor plant with WWER-1000
- Analyses of Beyond Design Basis Accident Homogeneous Boron Dilution Scenarios
- Analysis of heterogeneous boron dilution transients during outages with APROS 3D nodal core model
- Prospects of subcritical molten salt reactor for minor actinides incineration in closed fuel cycle
- Usage of burnt fuel isotopic compositions from engineering codes in Monte-Carlo code calculations
- Neutron-kinetic and thermo-hydraulic uncertainties in the study of Kalinin-3 benchmark
- Inter-assembly gap deviations in VVER-1000: Accounting for effects on engineering margin factors
