Some uncertainty results obtained by the statistical version of the KARATE code system related to core design and safety analysis
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I. Panka
Abstract
The best-estimate KARATE code system has been widely used for core design calculations and simulations of slow transients of VVER reactors. Recently there has been an increasing need for assessing the uncertainties of such calculations by propagating the basic input uncertainties of the models through the full calculation chain. In order to determine the uncertainties of quantities of interest during the burnup, the statistical version of the KARATE code system has been elaborated. In the first part of the paper, the main features of the new code system are discussed. The applied statistical method is based on Monte-Carlo sampling of the considered input data taking into account mainly the covariance matrices of the cross sections and/or the technological uncertainties. In the second part of the paper, only the uncertainties of cross sections are considered and an equilibrium cycle related to a VVER-440 type reactor is investigated. The burnup dependence of the uncertainties of some safety related parameters (e.g. critical boron concentration, rod worth, feedback coefficients, assembly-wise radial power and burnup distribution) are discussed and compared to the recently used limits.
Kurzfassung
Das Best-Estimate-KARATE-Codesystem wird für Kerndesignberechnungen und Simulationen von langsamen Transienten von VVER-Reaktoren verwendet. In jüngster Zeit gab es einen zunehmenden Bedarf, die Unsicherheiten dieser Berechnungen zu beurteilen, indem die Auswirkungen der grundlegenden Eingabeunsicherheiten der Modelle über die volle Berechnungskette bestimmt werden. Dazu wurde die statistische Version des KARATE-Codesystems erarbeitet. Im ersten Teil des Beitrags werden die Hauptmerkmale des neuen Codesystems diskutiert. Die angewendete statistische Methode basiert auf der Monte-Carlo-Bestimmung der zu betrachtenden Eingabedaten unter Berücksichtigung der Kovarianzmatrizen der Querschnitte und/oder der technologischen Unsicherheiten. Im zweiten Teil des Beitrags werden nur die Unsicherheiten der Querschnitte berücksichtigt und es wird ein Gleichgewichtszyklus eines Reaktors vom Typ VVER-440 untersucht. Die Abbrand-Abhängigkeiten der Unsicherheiten einiger sicherheitsrelevanter Parameter (z.B. kritische Borkonzentration, Stabwert, Rückkopplungskoeffizienten, montageweise radiale Leistung und Verbrennungsverteilung) werden diskutiert und mit den bislang verwendeten Grenzen verglichen.
References
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© 2017, Carl Hanser Verlag, München
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Modelling Human Resource Requirements for the Nuclear Industry in Europe
- Some uncertainty results obtained by the statistical version of the KARATE code system related to core design and safety analysis
- The integrity of NSSS and containment during extended station blackout for Kuosheng BWR plant
- Experimental investigation of effect of spacer on two phase turbulent mixing rate in subchannels of pressure tube type BWR
- Thermal-hydraulic analysis of research reactor core with different LEU fuel types using RELAP5
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- Dose assessment for emergency workers in early phase of Fukushima Daiichi nuclear power plant accident
- Anti-neutrino flux in a research reactor for non-proliferation application
- Friction coefficient and limiter load test analysis by flexibility coefficient model of Hold-Down Spring of nuclear reactor vessel internals
- Robust observer based control for axial offset in pressurized-water nuclear reactors based on the multipoint reactor model using Lyapunov approach
- Internal and external hazards inside the containment in case of an emergency situation
- Slab albedo for linearly and quadratically anisotropic scattering kernel with modified FN method
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Modelling Human Resource Requirements for the Nuclear Industry in Europe
- Some uncertainty results obtained by the statistical version of the KARATE code system related to core design and safety analysis
- The integrity of NSSS and containment during extended station blackout for Kuosheng BWR plant
- Experimental investigation of effect of spacer on two phase turbulent mixing rate in subchannels of pressure tube type BWR
- Thermal-hydraulic analysis of research reactor core with different LEU fuel types using RELAP5
- The application of knowledge management and TRIZ for solving the safe shutdown capability of fire alarms in nuclear power plants
- Dose assessment for emergency workers in early phase of Fukushima Daiichi nuclear power plant accident
- Anti-neutrino flux in a research reactor for non-proliferation application
- Friction coefficient and limiter load test analysis by flexibility coefficient model of Hold-Down Spring of nuclear reactor vessel internals
- Robust observer based control for axial offset in pressurized-water nuclear reactors based on the multipoint reactor model using Lyapunov approach
- Internal and external hazards inside the containment in case of an emergency situation
- Slab albedo for linearly and quadratically anisotropic scattering kernel with modified FN method
