Determination of the RBMK-1500 reactor passport characteristics
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R. Pabarcius
Abstract
During the whole life of reactor equipment it has to be shown that safety relevant limitations of general reactivity characteristics are not exceeded. Usually, safety margins of these characteristics are listed in a technical certificate (reactor passport). This paper presents the calculation results of the Ignalina nuclear power plant unit 2 reactor passport characteristics obtained by the best estimate 3D computer code QUABOX/CUBBOX, adjusted for the RBMK-1500 reactor core. The characteristics determined are supported by an uncertainty and sensitivity analysis which was performed using the software package SUSA. The uncertainty of the reactivity coefficients and associated effects were quantified in terms of statistical tolerance limits. The sensitivity analysis provided a ranking of the uncertain input parameters with respect to their contribution to the uncertainty of the reactivity characteristics.
Kurzfassung
Während der gesamten Lebensdauer von Reaktorkomponenten muss gezeigt werden, dass sicherheitsrelevante Grenzwerte in Bezug auf das Reaktivitätsverhalten nicht überschritten werden. In der Regel werden die Sicherheitsbereiche solcher Parameter in einer technischen Sicherheitsspezifikation festgelegt. Die vorliegende Arbeit präsentiert die Berechnungsergebnisse für die Sicherheitsspezifikationsparameter des Blockes 2 des Ignalina Kernkraftwerks unter Verwendung des 3D-Programms QUABOX/CUBBOX, angepasst an den RBMK-1500 Reaktorkern. Die Ergebnisse werden durch eine Unsicherheits- und Sensitivitätsanalyse mit Hilfe des Programmpakets SUSA ergänzt. Die Unsicherheit der Reaktivitätskoeffizienten und damit verbundener Effekte wurden in Form von statistischen Toleranzgrenzen quantifiziert. Die Sensitivitätsanalyse liefert eine Gewichtung der unsicheren Eingangsparameter nach ihrem Beitrag zur Unsicherheit bei der Bestimmung des Reaktivitätsverhaltens.
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© 2008, Carl Hanser Verlag, München
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Analysis of the hydrogen behaviour in compartments of the Ignalina nuclear power plant
- Determination of the RBMK-1500 reactor passport characteristics
- Technical feasibility of using RU-43 fuel in the CANDU-6 reactors of the Cernavoda NPP
- Computational study of moderator flow and temperature fields in the calandria vessel of a heavy water reactor using the PHOENICS code
- Regulatory requirements on level 2 PSA in Germany and their associated potential to improve emergency management
- Dynamic structure response due to reactor cooling piping system failure
- Uncertainty in activation cross-section calculations at intermediate proton energies
- The LTSN solution of the transport equation for one-dimensional cartesian geometry with c = 1
- Application of the UN approximation to the neutron transport equation in slab geometry
- The reflected critical slab problem for one-speed neutrons with strongly anisotropic scattering
- The analytical representation of the fundamental mode in 1-D-geometry for the CANDLE burn-up phenomenon
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Analysis of the hydrogen behaviour in compartments of the Ignalina nuclear power plant
- Determination of the RBMK-1500 reactor passport characteristics
- Technical feasibility of using RU-43 fuel in the CANDU-6 reactors of the Cernavoda NPP
- Computational study of moderator flow and temperature fields in the calandria vessel of a heavy water reactor using the PHOENICS code
- Regulatory requirements on level 2 PSA in Germany and their associated potential to improve emergency management
- Dynamic structure response due to reactor cooling piping system failure
- Uncertainty in activation cross-section calculations at intermediate proton energies
- The LTSN solution of the transport equation for one-dimensional cartesian geometry with c = 1
- Application of the UN approximation to the neutron transport equation in slab geometry
- The reflected critical slab problem for one-speed neutrons with strongly anisotropic scattering
- The analytical representation of the fundamental mode in 1-D-geometry for the CANDLE burn-up phenomenon
