Analysis of fuel rod behaviour during limiting RIA in RBMK plants
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R. Pabarcius
Abstract
Typically, reactivity initiated accidents for nuclear power plants with RBMK type reactors are classified as design basis accidents. However, for assessment of the real probability of these events, such kind of incidents should be considered more seriously and the same acceptance criteria as for operational occurrences should be applied to them. The spurious withdrawal of control rods, as limiting reactivity initiated accident (RIA), was analysed and results of the study are presented in this paper. The power distribution in fuel channels located around the affected control rods was calculated using the QUABOX/CUBBOX-HYCA code. Evaluating possible uncertainties it was found that linear power, i. e. heat generated per unit length of coolant channel, exceeds the acceptance criterion by a factor of 1.21. This paper presents the discussion how such a peak of power influences the behaviour of RBMK-1500 reactor fuel rods. For this purpose, the FEMAXI-6 code was applied. Results of the analysis show, that after increasing and subsequently decreasing of the linear power during an accident most parameters (pressure inside fuel rods, radial displacements, fuel clad gaps, etc.) return to their initial values and no phenomena occurred that would preclude the resumption of normal operation after termination of such an event.
Kurzfassung
Üblicherweise werden Reaktivitätsstörfälle in Kernkraftwerken mit Reaktoren vom Typ RBMK als Auslegungsstörfälle eingestuft. Für die Beurteilung der tatsächlichen Wahrscheinlichkeit solcher Ereignisse sollte diese Art von Störfällen jedoch ernster genommen und es sollten die gleichen Akzeptanzkriterien wie für betriebliche Ereignisse angewandt werden. Das unberechtigte Ausfahren von Steuerstäben, als limitierender Reaktivitätsstörfall wurde analysiert und Ergebnisse der Studie werden in diesem Beitrag vorgestellt. Die Leistungsverteilung im Brennstoffkanal rund um die betroffenen Steuerstäbe wurde mit Hilfe des QUABOX/CUBBOX-HYCA-Codes berechnet. Bei der Abschätzung möglicher Unsicherheiten stellte sich heraus, dass die lineare Leistung, das heißt die pro Längeneinheit des Kühlkanals erzeugte Wärme, die Akzeptanzkriterien um den Faktor 1,21 überschreitet. Dieser Beitrag beschreibt die Diskussion, wie ein derartiger Leistungspeak das Verhalten der Brennelemente in RBMK-1500-Reaktoren beeinflussen. Zu diesem Zweck wurde der FEMAXI-6-Code angewandt. Die Ergebnisse der Analyse zeigen, dass nach Erhöhung und anschließender Abnahme der linearen Leistung während eines Störfalls die meisten Parameter (Druck in den Brennelementen, radiale Verlagerungen, Spalten zwischen Brennelementhülle und Brennstoff usw.) zu den ursprünglichen Werten zurückkehren und keine Phänomene eintreten, die der Wiederaufnahme des normalen Betriebs nach Beendigung eines solchen Ereignisses entgegen stehen.
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© 2010, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries/Kurzfassungen
- Technical Contributions/Fachbeiträge
- Analyses of loads on reactor pressure vessel internals in a pressurized water reactor due to a loss-of-coolant accident considering fluid-structure interaction
- Preliminary evaluation of a radioactive waste repository safety performance by a Monte Carlo simulation-based reliability model
- Determination of effects of burn up and reflector material on the kinetic parameters for open pool reactor using MCNP code
- Analysis of fuel rod behaviour during limiting RIA in RBMK plants
- Reflection on the ductility of irradiated zircaloy-4 fuel rod cladding
- Prediction, analysis and solution of flow inversion phenomenon in a typical MTR reactor with upward core cooling
- Steam drum process dynamics and level control of a pressure tube BWR
- Nuclear data for cyclotron production of 114mIn/114In and 140Nd/140Pr used in gamma camera monitoring, RIT, ERT and PET
- Effect of thermal gap conductance for MoO3 ampoules irradiated in a high neutron flux
- U1 approximation to the neutron transport equation and calculation of the asymptotic relaxation length
- Application of the TN method to critical slab problem for one-speed neutrons with forward and backward scattering and efficiency of reflection coefficient
- Technical Notes/Technische Mitteilungen
- Rapid preparation of Uranium and Thorium alpha sources by electroplating technique
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries/Kurzfassungen
- Technical Contributions/Fachbeiträge
- Analyses of loads on reactor pressure vessel internals in a pressurized water reactor due to a loss-of-coolant accident considering fluid-structure interaction
- Preliminary evaluation of a radioactive waste repository safety performance by a Monte Carlo simulation-based reliability model
- Determination of effects of burn up and reflector material on the kinetic parameters for open pool reactor using MCNP code
- Analysis of fuel rod behaviour during limiting RIA in RBMK plants
- Reflection on the ductility of irradiated zircaloy-4 fuel rod cladding
- Prediction, analysis and solution of flow inversion phenomenon in a typical MTR reactor with upward core cooling
- Steam drum process dynamics and level control of a pressure tube BWR
- Nuclear data for cyclotron production of 114mIn/114In and 140Nd/140Pr used in gamma camera monitoring, RIT, ERT and PET
- Effect of thermal gap conductance for MoO3 ampoules irradiated in a high neutron flux
- U1 approximation to the neutron transport equation and calculation of the asymptotic relaxation length
- Application of the TN method to critical slab problem for one-speed neutrons with forward and backward scattering and efficiency of reflection coefficient
- Technical Notes/Technische Mitteilungen
- Rapid preparation of Uranium and Thorium alpha sources by electroplating technique
