Coupled neutronic core and subchannel analysis of nanofluids in VVER-1000 type reactor
-
E. Zarifi
,
G. Jahanfarnia
Abstract
This study is aimed to perform the coupled thermal-hydraulic/neutronic analysis of nanofluids as the coolant in the hot fuel assembly of VVER-1000 reactor core. Water-based nanofluid containing various volume fractions of Al2O3 nanoparticle is analyzed. WIMS and CITATION codes are used for neutronic simulation of the reactor core, calculating neutron flux and thermal power distribution. In the thermal-hydraulic modeling, the porous media approach is used to analyze the thermal behavior of the reactor core and the subchannel analysis is used to calculate the hottest fuel assembly thermal-hydraulic parameters. The derived conservation equations for coolant and conduction heat transfer equation for fuel and clad are discretized by Finite volume method and solved numerically using visual FORTRAN program. Finally the analysis results for nanofluids and pure water are compared together. The achieved results show that at low concentration (0.1 percent volume fraction) alumina is the optimum nanoparticles for normal reactor operation.
Abstract
In diesem Beitrag wird eine gekoppelte neutronenphysikalische und thermodynamische Analyse von Nanofluiden als Kühlmittel in der Brennelement-Anordnung eines VVER-Reaktorkerns durchgeführt. Nanofluide auf Wasser-Basis, die verschiedene Volumenanteile von Al2O3 Nanopartikeln enthalten, wurden analysiert. WIMS und CITATION-Codes wurden für die neutronenphysikalische Simulation des Reaktorkerns verwendet. Bei der thermohydraulischen Modellierung wurde der Porous-Media-Ansatz zur Analyse des Wärmeverhaltens des Reaktorkerns verwendet und die Unterkanalanalyse wurde zur Berechnung der thermohydraulischen Parameter der heißesten Brennelement-Anordnung verwendet. Erhaltungsgleichungen und Wärmeübertragungseigenschaften wurden für Brennelemente und Hülle abgeleitet und mit Hilfe der Finite-Volumen-Methode diskretisiert und numerisch gelöst. Die Ergebnisse der Nanofluid-Analyse wurden verglichen mit den Ergebnissen für reines Wasser. Es zeigt sich, dass bei niedrigen Konzentrationen (0.1% Volumenanteile) Aluminiumoxid die optimalen Nanopartikel für den normalen Reaktorbetrieb sind.
Nomenclature
- A
flow area [m2]
- C
specific heat capacity [J/kgK]
- cba
critical boron concentration in the primary coolant loop [g/kg]
- d
nanoparticle diameter [nm]
- i
enthalpy [J/kg]
- g
gravitational acceleration [m/s2]
- h
convective heat transfer coefficient [w/K]
- H10
location of control rods of type 10 [%]
- K
thermal conductivity, [w/mK]
- kB
Boltzman constant (1.3807 × 10–23) [J/K]
- kG
pressure drop coefficient
- l
Center to center distance between two adjacent channels [m]
- N
reactor thermal power [Mw]
- Nu
Nusselt number
- P
pressure [Pa]
- Pr
Prandtl number
- q′
linear heat rate [w/m]
- q′′
heat flux [w/m2]
- q′′′
volumetric heat generation [w/m3]
- r
radius[m]
- Rem
modified Reynolds number
- Re
Reynolds number
- Sk
lateral length in “k” gap [m]
- sl
lateral closed physical area [m2]
- T
temperature [K]
- teff
effective time [day]
- T
coolant temperature [8C]
- u
internal energy [J/kg]
- v
velocity [m/s]
- w
mass flow [kg/s]
Greek symbols
- α
molecular thermal diffusivity [m2/s]
- γ
porosity
- μ
viscosity [Nsm–1]
- ρ
density [kg/m3]
- φ
volume fraction
- Φ
dissipation function [w/m3]
- Ψ
correction factor
- τ
Shear stress [Pa]
Subscripts
- b
coolant bulk
- ci
clad inner surface
- co
clad outer surface
- c.t.
circular tube
- f
base fluid
- fo
fuel outer surface
- g
gas
- l, l0
channel Number
- nf
nanofluids
- p
nanoparticle
- w
wall surface
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Books · Bücher
Uranium 2016: Resources, Production and Demand. NEA report No. 7301, Published by OECD/NEA, 2016, 548 pp., in English, available online at: http://www.oecd-nea.org/ndd/pubs/2016/7301-uranium-2016.pdf
Uranium is the raw material used to produce fuel for long-lived nuclear power facilities, necessary for the generation of significant amounts of baseload low-carbon electricity for decades to come. Although a valuable commodity, declining market prices for uranium in recent years, driven by uncertainties concerning evolutions in the use of nuclear power, have led to the postponement of mine development plans in a number of countries and to some questions being raised about future uranium supply. This 26th edition of the “Red Book”, a recognised world reference on uranium jointly prepared by the Nuclear Energy Agency (NEA) and the International Atomic Energy Agency (IAEA), provides analyses and information from 49 producing and consuming countries in order to address these and other questions. The present edition provides the most recent review of world uranium market fundamentals and presents data on global uranium exploration, resources, production and reactor-related requirements. It offers updated information on established uranium production centres and mine development plans, as well as projections of nuclear generating capacity and reactor-related requirements through 2035, in order to address long-term uranium supply and demand issues.
© 2017 Carl Hanser Verlag GmbH & Co. KG
Artikel in diesem Heft
- Frontmatter
- Best estimate approach for the evaluation of critical heat flux phenomenon in the boiling water reactors
- Coupled neutronic core and subchannel analysis of nanofluids in VVER-1000 type reactor
- PCTRAN enhancement for large break loss of coolant accident concurrent with loss of offsite power in VVER-1000 simulation
- Experimental study of natural circulation flow instability in rectangular channels
- Exerimental method and preliminary studies of the passive containment water film evaporation mass transfer
- Automated generation of burnup chain for reactor analysis applications
- Fission source sampling in coupled Monte Carlo simulations
- Hysteresis phenomenon in nuclear reactor dynamics
- Investigation of neutronic and safety parameters variation in 5 MW research reactor due to U3O8Al fuel conversion to ThO2 + U3O8Al
- Implementation of meso-scale radioactive dispersion model for GPU
- Solution of the multilayer multigroup neutron diffusion equation in cartesian geometry by fictitious borders power method
- Half-space albedo problem with modified FN method for linear and quadratic anisotropic scattering
Artikel in diesem Heft
- Frontmatter
- Best estimate approach for the evaluation of critical heat flux phenomenon in the boiling water reactors
- Coupled neutronic core and subchannel analysis of nanofluids in VVER-1000 type reactor
- PCTRAN enhancement for large break loss of coolant accident concurrent with loss of offsite power in VVER-1000 simulation
- Experimental study of natural circulation flow instability in rectangular channels
- Exerimental method and preliminary studies of the passive containment water film evaporation mass transfer
- Automated generation of burnup chain for reactor analysis applications
- Fission source sampling in coupled Monte Carlo simulations
- Hysteresis phenomenon in nuclear reactor dynamics
- Investigation of neutronic and safety parameters variation in 5 MW research reactor due to U3O8Al fuel conversion to ThO2 + U3O8Al
- Implementation of meso-scale radioactive dispersion model for GPU
- Solution of the multilayer multigroup neutron diffusion equation in cartesian geometry by fictitious borders power method
- Half-space albedo problem with modified FN method for linear and quadratic anisotropic scattering
