Simulation of turbulent mixing rate in simulated subchannels of a reactor rod bundle
-
M. P. Sharma
and
A. Moharana
Abstract
Subchannel analysis codes are widely used for the thermal-hydraulic design of nuclear reactor rod bundle. The effectiveness of subchannel analysis codes depends on turbulent mixing between these subchannels. Turbulent mixing has no direct contribution to the axial mass flow rate through subchannel but it will cause exchange of momentum and energy between the neighboring subchannels. Thus, it is important to evaluate the turbulent mixing coefficient for reactor rod bundle as it is a significant factor in the lateral energy and momentum equation for subchannel analysis codes like COBRA IIIC, COBRA-IV and MATRA LMR-FB.
With the rapid developments in computational fluid dynamics and computer performance, three-dimensional analyses of turbulent flows occurring in the nuclear rod bundle have become more prominent. Several numerical analyses have already been attempted to investigate the flow behavior in rod bundles of different reactors. Much of these are dedicated to find out the structure of turbulence in rod bundle but a few analyses has been done to evaluate the magnitude of the turbulent mixing coefficient. In view of this, CFD analyses were carried out to determine the turbulent mixing coefficient in the simulated sub-channels of the reactor rod bundle. Previous studies on the structure of turbulence reveals that it is highly anisotropic. Hence, the Reynolds Stress Model (RSM), finer mesh and near wall distance ( y + ≤ 2) is required to capture turbulent mixing phenomena. The validation of results is done by comparing with subchannel mixing experiments.
Abstract
Unterkanal-Analysecodes werden für die thermohydraulische Auslegung von Brennstabbündeln verwendet. Dabei muss der Analysecode die turbulente Vermischung in den Unterkanälen korrekt abbilden. Die turbulente Vermischung hat keinen direkten Beitrag zum axialen Massenstrom durch den Unterkanal, bewirkt jedoch einen Impuls- und Energieaustausch zwischen den benachbarten Unterkanälen. Daher ist es wichtig, den turbulenten Mischungskoeffizienten für das Brennstabbündel zu bewerten, da er ein wesentlicher Faktor in der lateralen Energie- und Impulsgleichung für Subkanalanalysecodes wie COBRA IIIC, COBRA-IV und MATRA LMR-FB ist.
Mit den rasanten Entwicklungen in der rechnergestützten Fluiddynamik und der Computerleistung sind dreidimensionale Analysen turbulenter Strömungen im Kernstabbündel stärker in den Vordergrund gerückt. Es wurden bereits mehrere numerische Analysen durchgeführt, um das Fließverhalten in Stabbündeln verschiedener Reaktoren zu untersuchen. Viele davon widmen sich der Ermittlung der Turbulenzstruktur im Stabbündel. Es wurden auch einige Analysen durchgeführt, um die Größe des turbulenten Mischungskoeffizienten zu bewerten. In diesem Beitrag werden CFD-Analysen beschrieben, bei denen turbulente Mischungskoeffizienten in den simulierten Unterkanälen des Brennstabbündels bestimmt werden. Frühere Studien zur Struktur von Turbulenzen haben gezeigt, dass diese stark anisotrop sind. Daher ist das Reynolds-Spannungsmodell (RSM), ein feineres Netz und ein wandnaher Abstand ( y + ≤ 2) erforderlich, um turbulente Mischungsphänomene zu erfassen. Die Validierung der Ergebnisse erfolgt durch Vergleich mit ausgewählten Experimenten.
Nomenclature
- A
flow area (m2)
- Dh
hydraulic diameter (m)
- d
rod diameter (m)
- Re
Reynolds number
- S
gap (m)
- t
time
- UU
Reynold stress in x
- VV
Reynold stress in y
- WW
Reynold stress in z
- W’
turbulent mixing rate (Kg/m-s)
Subscript
- i,j,k
subchannel identifier
- 1,2,3
subchannel identifier
- cfd
computation fluid dynamics
- exp
experiment
- avg
average
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© 2021 Walter de Gruyter GmbH, Berlin/Boston, Germany
Articles in the same Issue
- Frontmatter
- Study of PWR hot leg creep rupture and RCS depressurization strategy during an SBO accident
- Effect of gap design pressure on the LWR fuel rods lifetime
- Simulation of turbulent mixing rate in simulated subchannels of a reactor rod bundle
- Methodology for analyzing accidents with radioactive material release with code EPZDose
- Development and usage of the digital SAMG system
- Analysis of Doppler reactivity of SMART reactor core for hybrid fuel configurations of UO2, MOX and (Th/U)O2 using OpenMC
- Corrosion surveillance program for tank, fuel cladding and supporting structure of 30 MW Indonesian RSG GAS research reactor
- Investigation of effects of nonavailability of passive safety systems on the reactor behaviour during LOCA Scenario in AP600
Articles in the same Issue
- Frontmatter
- Study of PWR hot leg creep rupture and RCS depressurization strategy during an SBO accident
- Effect of gap design pressure on the LWR fuel rods lifetime
- Simulation of turbulent mixing rate in simulated subchannels of a reactor rod bundle
- Methodology for analyzing accidents with radioactive material release with code EPZDose
- Development and usage of the digital SAMG system
- Analysis of Doppler reactivity of SMART reactor core for hybrid fuel configurations of UO2, MOX and (Th/U)O2 using OpenMC
- Corrosion surveillance program for tank, fuel cladding and supporting structure of 30 MW Indonesian RSG GAS research reactor
- Investigation of effects of nonavailability of passive safety systems on the reactor behaviour during LOCA Scenario in AP600
