Steady state and transient analyses of MNSR reactor using RELAP5 code
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Abstract
Developing a reliable thermal-hydraulic model of a nuclear reactor is an essential process in the steady state and transient analyses. This paper provides the results of best estimate calculation carried out with reference to Iranian Miniature Neutron Source Reactor (MNSR) using the RELAP5 code. Applying the qualified nodalization and the cross-flow effects are some of the advantages in the present model. Here, various transients including step and ramp reactivity insertions were inspected for safety analysis. The obtained results from the code showed a reasonable agreement with the MNSR Safety Analysis Report (SAR) and existing experimental and reference data.
Kurzfassung
Die Entwicklung eines zuverlässigen thermohydraulischen Modells eines Kernreaktors ist ein wichtiger Schritt bei der Analyse des stationären und transienten Verhaltens eines Reaktors. Dieser Beitrag stellt die Ergebnisse der Best-Estimate-Berechnungen vor, die für den iranischen Miniatur-Neutronenquelle-Reaktor (MNSR) mit Hilfe des RELAP5-Codes durchgeführt wurden. Die Anwendung qualifizierter Nodalisation und Querströmungseffekte bieten einige Vorteile im vorliegenden Modell. Verschiedene Transienten einschließlich des schrittweisen Einbringens der Reaktivität wurden für die Sicherheitsanalyse untersucht. Die erhaltenen Ergebnisse stimmen überein mit den Daten des MNSR-Berichts zur Sicherheitsanalyse (SAR) und vorhandenen experimentellen Daten.
References
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© 2016, Carl Hanser Verlag, München
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Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Stability analysis of the Korean prototype Generation-IV sodium-cooled fast reactor using linear frequency domain approach
- Validation of RELAP5 model of experimental test rig simulating the natural convection in MTR research reactors
- Steady state and transient analyses of MNSR reactor using RELAP5 code
- Protactinium-231 as a new fissionable material for nuclear reactors that can produce nuclear fuel with stable neutron-multiplying properties
- Assessment of pin-by-pin fission rate distribution within MOX/UO2 fuel assembly using MCNPX code
- Influence on rewetting temperature and wetting delay during rewetting rod bundle by various radial jet models
- Experimental and numerical investigation on natural convection heat transfer in nanofluids
- Experimental studies in a single-phase parallel channel natural circulation system: preliminary results
- Calculation of PDS-XADS core closed-loop transfer function by using feedback with the lumped-model
- Calculation of nuclear reactivity using the generalised Adams-Bashforth-Moulton predictor corrector method
