Calculation of the fuel temperature coefficient of reactivity considering non-uniform radial temperature distribution in the fuel rod
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A. Pazirandeh
Abstract
The safe operation of a reactor is based on feedback models. In this paper we attempted to discuss the influence of a non-uniform radial temperature distribution on the fuel rod temperature coefficient of reactivity. The paper demonstrates that the neutron properties of a reactor core is based on effective temperature of the fuel to obtain the correct fuel temperature feedback. The value of volume-averaged temperature being used in the calculations of neutron physics with feedbacks would result in underestimating the probable event. In the calculation it is necessary to use the effective temperature of the fuel in order to provide correct accounting of the fuel temperature feedback. Fuel temperature changes in different zones of the core and consequently reactivity coefficient change are an important parameter for analysis of transient conditions. The restricting factor that compensates the inserted reactivity is the temperature reactivity coefficient and effective delayed neutron fraction.
Kurzfassung
Der sichere Betrieb von Reaktoren basiert auf Rückmeldungsmodellen. In diesem Beitrag wird versucht, den Einfluss einer ungleichförmigen radialen Temperaturverteilung auf den Temperaturkoeffizienten der Reaktivität der Brennelemente zu diskutieren. Es wird gezeigt, dass die Eigenschaften der Neutronen des Reaktorkerns auf der effektiven Temperatur des Brennstoffs basieren, um die korrekte Rückmeldung der Brennelementtemperatur zu erhalten. Der Wert der in den neutronenphysikalischen Berechnungen verwendeten volumen-gemittelten Temperatur mit Rückmeldungen würde zu einer Unterschätzung des möglichen Ereignisses führen. Es ist deshalb nötig, in den Berechnungen die effektive Temperatur des Brennstoffs zu verwenden, um so die korrekte Berechnung für die Brennstofftemperatur zu erhalten. Änderungen der Brennstofftemperatur in verschiedenen Zonen des Kerns und damit Änderungen des Reaktivitätskoeffizienten sind wichtige Parameter zur Analyse transienter Bedingungen. Der einschränkende Faktor, der die eingesetzte Reaktivität kompensiert, ist der Temperatur-Reaktivitätskoeffizient und der Anteil effektiv verzögerter Neutronen.
References
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© 2017, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- SIMULATE-3 K coupled code applications
- Application of the new IAPWS Guideline on the fast and accurate calculation of steam and water properties with the Spline-Based Table Look-Up Method (SBTL) in RELAP-7
- Simulation of water hammer phenomena using the system code ATHLET
- New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses
- Sensitivity analysis for thermo-hydraulics model of a Westinghouse type PWR: verification of the simulation results
- Calculation of the fuel temperature coefficient of reactivity considering non-uniform radial temperature distribution in the fuel rod
- The effect of boron dilution transient on the VVER-1000 reactor core using MCNP and COBRA-EN codes
- Modelling of the spent fuel heat-up in the spent fuel pools using one-dimensional system codes and CFD codes
- Optimization and analysis of the effects of physical parameters in a TRIGA-ADSR
- A comparison study for mass attenuation coefficients of some amino acids using MCNP code
- Validation of radioactive isotope activity measurement in homogeneous waste drum using Monte Carlo codes
- Study of the response reduction of LiF:Mg, Ti dosimeter for high dose dosimetry
- Non-contact micro mass evaluation method using an X-ray microscope
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- SIMULATE-3 K coupled code applications
- Application of the new IAPWS Guideline on the fast and accurate calculation of steam and water properties with the Spline-Based Table Look-Up Method (SBTL) in RELAP-7
- Simulation of water hammer phenomena using the system code ATHLET
- New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses
- Sensitivity analysis for thermo-hydraulics model of a Westinghouse type PWR: verification of the simulation results
- Calculation of the fuel temperature coefficient of reactivity considering non-uniform radial temperature distribution in the fuel rod
- The effect of boron dilution transient on the VVER-1000 reactor core using MCNP and COBRA-EN codes
- Modelling of the spent fuel heat-up in the spent fuel pools using one-dimensional system codes and CFD codes
- Optimization and analysis of the effects of physical parameters in a TRIGA-ADSR
- A comparison study for mass attenuation coefficients of some amino acids using MCNP code
- Validation of radioactive isotope activity measurement in homogeneous waste drum using Monte Carlo codes
- Study of the response reduction of LiF:Mg, Ti dosimeter for high dose dosimetry
- Non-contact micro mass evaluation method using an X-ray microscope
