Assessment of the uncertainties of MULTICELL calculations by the OECD NEA UAM PWR pin cell burnup benchmark
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I. Panka
Abstract
Defining precisely the burnup of the nuclear fuel is important from the point of view of core design calculations, safety analyses, criticality calculations (e. g. burnup credit calculations), etc. This paper deals with the uncertainties of MULTICELL calculations obtained by the solution of the OECD NEA UAM PWR pin cell burnup benchmark. In this assessment Monte-Carlo type statistical analyses are applied and the energy dependent covariance matrices of the cross-sections are taken into account. Additionally, the impact of the uncertainties of the fission yields is also considered. The target quantities are the burnup dependent uncertainties of the infinite multiplication factor, the two-group cross-sections, the reaction rates and the number densities of some isotopes up to the burnup of 60 MWd/kgU. In the paper the burnup dependent tendencies of the corresponding uncertainties and their sources are analyzed.
Kurzfassung
Für Kernauslegungsrechnungen, Sicherheitsanalysen und Kritikalitätsuntersuchungen ist die genaue Bestimmung des Abbrands des nuklearen Brennstoffs eine wichtige Grundvoraussetzung. In diesem Beitrag werden die Ungenauigkeiten der Ergebnisse der Rechnungen des OECD NEA UAM PWR pin cell burnup Benchmarks mit dem Programm MULTICELL bestimmt. Dabei wurden statistische Monte-Carlo Analysen angewendet sowie energieabhängige Kovarianzmatrizen der Querschnittswerte und der Einfluss der Unsicherheiten der Spaltausbeute berücksichtigt. Zielgrößen waren vom Abbrand abhängige Unsicherheiten des unendlichen Multiplikationsfaktors, der Zweigruppenquerschnitte, der Reaktionsraten und der Dichten von Isotopen bis zu einem Abbrand von 60 MWd/kgU. Die vom Abbrand abhängigen Tendenzen der entsprechenden Unsicherheiten und deren Ursachen werden analysiert.
References
1 Ivanov, K.; Avramova, M.; Kamerow, S.; Kodeli, I.; Sartori, E.; Ivanov, E.; Cabellos, O.: Benchmark for uncertainty analysis in modeling (UAM) for design, operation and safety analysis of LWRs, Volume I: Specification and Support Data for the Neutronics Cases (Phase I), Version 2.1 (Final Specifications), May 2013, NEA/NSC/DOC (2013) 7Search in Google Scholar
2 Keresztúri, A.; Hegyi, Gy.; Hordósy, G.; Makai, M.; Maráczy, Cs.; Telbisz, M.: KARATE – A Code System for VVER-440 Core Calculations. Proceedings of the 5th Symposium of AER, Dobogókő, 1995Search in Google Scholar
3 Hegyi, Gy.; Temesvári, E.; Keresztúri, A.; Maráczy, Cs.; Makai, M.: Upgrading of the Core calculation Modules of KARATE-440. Proceedings of the 9th Symposium of AER, Demanovska Dolina, Slovakia, October 4–8, 1999Search in Google Scholar
4 Broadhead;B. L.: SCALE 5.1 Cross-Section Covariance Libraries, ORNL/TM-2005/39,Version 5.1, Vol. I, Book 3, Sect. M19, November 2006Search in Google Scholar
5 OECD NEA JANIS database, http://www.oecd-nea.org/janis/Search in Google Scholar
6 Martinez, J. S.; Zwermann, W.; Gallner, L.; Puente-Espel, F.; Cabellos, O.; Velkov, K.; Hannstein, V.: Propagation of Neutron Cross-section, Fission Yield, and Decay Data Uncertainties in Depletion Calculations, Nuclear Data Sheets118 (2014) 480–48310.1016/j.nds.2014.04.112Search in Google Scholar
7 Cabellos, O.: Presentation and Discussion of the UAM/Exercise I-1b: “Pin-Cell Burnup Benchmark” with the Hybrid Method, Science and Technology of Nuclear Installations, Volume 2013 (2013), Article ID 790206, 12 pages 10.1155/2013/790206Search in Google Scholar
© 2015, Carl Hanser Verlag, München
Articles in the same Issue
- 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
Articles in the same Issue
- 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
