Sensitivity of the neutron multiplication factor to water ingress into a spent fuel cask
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M. J. Leotlela
Abstract
An essential component of safety analyses is the investigation of accident scenarios. In this paper water ingress scenarios of spent fuel containers, as they may occur during transport or storage, are examined. In the main body of this paper, a number of paths are studied through which water can gain access to the spent fuel cask and eventually reach the fuel pellet, potentially resulting in an increase in reactivity as a result of over-moderation. The primary objective of this project was to perform an assessment of what, in the unlikely event of a Fukushima- type accident, the impact would be on the reactivity of the cask by analyzing a gradual increase in water level in the spent fuel casks. In addition, the way the keff of the system responds to such an increase is discussed. The paper also provides the results of an assessment of the reactivity effect of water ingress via various pathways/channels.
© 2020 by Walter de Gruyter Berlin/Boston
Articles in the same Issue
- CONTENTS
- CALENDAR OF EVENTS · VERANSTALTUNGSKALENDER
- Verification of the stationary module of the ShIPR software system for modelling experiments of the ASTRA HTGR type critical facility
- Appraisal of new density coefficient on integrated-nanoparticles concrete in nuclear protection
- Comparison of back propagation network and radial basis function network in Departure from Nucleate Boiling Ratio (DNBR) calculation
- Developing a futuristic multi-objective optimization of the fuel management problems for the nuclear research reactors
- Sensitivity of the neutron multiplication factor to water ingress into a spent fuel cask
- Evaluation of single heated channel and subchannel modeling of a nuclear once through steam generator (OTSG)
- Note · Imprint
Articles in the same Issue
- CONTENTS
- CALENDAR OF EVENTS · VERANSTALTUNGSKALENDER
- Verification of the stationary module of the ShIPR software system for modelling experiments of the ASTRA HTGR type critical facility
- Appraisal of new density coefficient on integrated-nanoparticles concrete in nuclear protection
- Comparison of back propagation network and radial basis function network in Departure from Nucleate Boiling Ratio (DNBR) calculation
- Developing a futuristic multi-objective optimization of the fuel management problems for the nuclear research reactors
- Sensitivity of the neutron multiplication factor to water ingress into a spent fuel cask
- Evaluation of single heated channel and subchannel modeling of a nuclear once through steam generator (OTSG)
- Note · Imprint
