A Coupled Model for Two-Phase Simulation of a Heavy Water Pressure Vessel Reactor
A Multi-dimensional computational fluid dynamics
(CFD) two-phase model was developed with the aim to simulate
the in-core coolant circuit of a pressurized heavy water reactor
(PHWR) of a commercial nuclear power plant (NPP). Due to the
fact that this PHWR is a Reactor Pressure Vessel type (RPV),
three-dimensional (3D) detailed modelling of the large reservoirs of
the RPV (the upper and lower plenums and the downcomer) were
coupled with an in-house finite volume one-dimensional (1D) code
in order to model the 451 coolant channels housing the nuclear fuel.
Regarding the 1D code, suitable empirical correlations for taking into
account the in-channel distributed (friction losses) and concentrated
(spacer grids, inlet and outlet throttles) pressure losses were used.
A local power distribution at each one of the coolant channels
was also taken into account. The heat transfer between the coolant
and the surrounding moderator was accurately calculated using a
two-dimensional theoretical model. The implementation of subcooled
boiling and condensation models in the 1D code along with the use
of functions for representing the thermal and dynamic properties of
the coolant and moderator (heavy water) allow to have estimations
of the in-core steam generation under nominal flow conditions for a
generic fission power distribution. The in-core mass flow distribution
results for steady state nominal conditions are in agreement with the
expected from design, thus getting a first assessment of the coupled
1/3D model. Results for nominal condition were compared with
those obtained with a previous 1/3D single-phase model getting more
realistic temperature patterns, also allowing visualize low values of
void fraction inside the upper plenum. It must be mentioned that the
current results were obtained by imposing prescribed fission power
functions from literature. Therefore, results are showed with the aim
of point out the potentiality of the developed model.
[1] Adorni, M., and Del Nevo, A., and DA´ uria, Francesco and Mazzantini,
OscarA procedure to address the fuel rod failures during LB-LOCA
transient in Atucha-2 NPP
[2] O. Mazzantini, M. Schivo,J. Csare, R. Garbero, M. Rivero and G. Theler.
A coupled calculation suite for atucha ii operational transients analysis,
Science and Technology of Nuclear Installations, 2010.
[3] M. Pecchia, Application of MCNP for Predicting Power Excursion During
LOCA in Atucha-2 PHWR, PhD Thesis, Univ. di Pisa, 2012.
[4] L. Vyskocil and J. Macek, Coupling CFD code with system code and
neutron kinetic code, Nuclear Engineering and Design, Vol 279, 210-218,
2014.
[5] D. Aumiller, E. Tomlinson, W. Weaver, An integrated relap5-3d and
multiphase CFD code system utilizing a semi-implicit coupling technique.
Nuclear engineering and design, 216(1), 77-87, 2002.
[6] D. Ramajo, S. Corzo,N. Schiliuk and N. Nigro. 3d modeling of the
primary circuit in the reactor pressure vessel of a phwr, Nuclear
Engineering and Design, Vol 265, 356-365, 2013.
[7] S. Corzo,D. Ramajo and N. Nigro. 1/3d modeling of the core coolant
circuit of a phwr nuclear power plant, Annals of Nuclear Energy, Vol 83,
386-397, 2015.
[8] D. Ramajo, S. Corzo, N. Schiliuk, A. Lazarte and N. Nigro. CFD
Modeling of the Moderator Tank of a PHWR Nuclear Power Plant, ENIEF
2014, Vol XXXIII, 2913-2926, 2014.
[9] S. Corzo. Assessment of Nuclear Power Reactor using Computational
Fluid Dynamics., PhD Thesis. Univ. Nac. del Litoral, 2015.
[10] J. Ferziger and M. Peric, Comp. Methods for Fluid Dynamics, Vol 3,
Springer, berlin, 1999.
[11] R. Lahey. A mechanistic subcooled boiling model. Proceedings of the
6th International Heat Transfer Conference, vol. 1, pp. 293-297, 1978.
[12] H. Unal. Maximum bubble diameter, maximum bubble-growth time and
bubble-growth rate during the subcooled nucleate flow boiling of water
up to 17.7mn/sq m. International Journal of Heat and Mass Transfer 19,
643-649, 1976.
[13] H. Unal. Determination of the initial point of net vapor generation in
flow boiling systems. International Journal of Heat and Mass Transfer
18(9), 1095-1099, 1975.
[14] S. Rouhani, E. Axelsson. Calculation of void volume fraction in the
subcooled and quality boiling regions. International Journal of Heat and
Mass Transfer, 13(2), 383-393, 1970.
[15] J. Chiang, B. Pei, F. Tsai, Pressurized water reactor (PWR) hot-leg
streaming: Part 1: Computational fluid dynamics (C) simulations. Nuclear
Engineering and Design 241(5), 1768-1775, 2011.
[1] Adorni, M., and Del Nevo, A., and DA´ uria, Francesco and Mazzantini,
OscarA procedure to address the fuel rod failures during LB-LOCA
transient in Atucha-2 NPP
[2] O. Mazzantini, M. Schivo,J. Csare, R. Garbero, M. Rivero and G. Theler.
A coupled calculation suite for atucha ii operational transients analysis,
Science and Technology of Nuclear Installations, 2010.
[3] M. Pecchia, Application of MCNP for Predicting Power Excursion During
LOCA in Atucha-2 PHWR, PhD Thesis, Univ. di Pisa, 2012.
[4] L. Vyskocil and J. Macek, Coupling CFD code with system code and
neutron kinetic code, Nuclear Engineering and Design, Vol 279, 210-218,
2014.
[5] D. Aumiller, E. Tomlinson, W. Weaver, An integrated relap5-3d and
multiphase CFD code system utilizing a semi-implicit coupling technique.
Nuclear engineering and design, 216(1), 77-87, 2002.
[6] D. Ramajo, S. Corzo,N. Schiliuk and N. Nigro. 3d modeling of the
primary circuit in the reactor pressure vessel of a phwr, Nuclear
Engineering and Design, Vol 265, 356-365, 2013.
[7] S. Corzo,D. Ramajo and N. Nigro. 1/3d modeling of the core coolant
circuit of a phwr nuclear power plant, Annals of Nuclear Energy, Vol 83,
386-397, 2015.
[8] D. Ramajo, S. Corzo, N. Schiliuk, A. Lazarte and N. Nigro. CFD
Modeling of the Moderator Tank of a PHWR Nuclear Power Plant, ENIEF
2014, Vol XXXIII, 2913-2926, 2014.
[9] S. Corzo. Assessment of Nuclear Power Reactor using Computational
Fluid Dynamics., PhD Thesis. Univ. Nac. del Litoral, 2015.
[10] J. Ferziger and M. Peric, Comp. Methods for Fluid Dynamics, Vol 3,
Springer, berlin, 1999.
[11] R. Lahey. A mechanistic subcooled boiling model. Proceedings of the
6th International Heat Transfer Conference, vol. 1, pp. 293-297, 1978.
[12] H. Unal. Maximum bubble diameter, maximum bubble-growth time and
bubble-growth rate during the subcooled nucleate flow boiling of water
up to 17.7mn/sq m. International Journal of Heat and Mass Transfer 19,
643-649, 1976.
[13] H. Unal. Determination of the initial point of net vapor generation in
flow boiling systems. International Journal of Heat and Mass Transfer
18(9), 1095-1099, 1975.
[14] S. Rouhani, E. Axelsson. Calculation of void volume fraction in the
subcooled and quality boiling regions. International Journal of Heat and
Mass Transfer, 13(2), 383-393, 1970.
[15] J. Chiang, B. Pei, F. Tsai, Pressurized water reactor (PWR) hot-leg
streaming: Part 1: Computational fluid dynamics (C) simulations. Nuclear
Engineering and Design 241(5), 1768-1775, 2011.
@article{"International Journal of Engineering, Mathematical and Physical Sciences:71393", author = "Damian Ramajo and Santiago Corzo and Norberto Nigro", title = "A Coupled Model for Two-Phase Simulation of a Heavy Water Pressure Vessel Reactor", abstract = "A Multi-dimensional computational fluid dynamics
(CFD) two-phase model was developed with the aim to simulate
the in-core coolant circuit of a pressurized heavy water reactor
(PHWR) of a commercial nuclear power plant (NPP). Due to the
fact that this PHWR is a Reactor Pressure Vessel type (RPV),
three-dimensional (3D) detailed modelling of the large reservoirs of
the RPV (the upper and lower plenums and the downcomer) were
coupled with an in-house finite volume one-dimensional (1D) code
in order to model the 451 coolant channels housing the nuclear fuel.
Regarding the 1D code, suitable empirical correlations for taking into
account the in-channel distributed (friction losses) and concentrated
(spacer grids, inlet and outlet throttles) pressure losses were used.
A local power distribution at each one of the coolant channels
was also taken into account. The heat transfer between the coolant
and the surrounding moderator was accurately calculated using a
two-dimensional theoretical model. The implementation of subcooled
boiling and condensation models in the 1D code along with the use
of functions for representing the thermal and dynamic properties of
the coolant and moderator (heavy water) allow to have estimations
of the in-core steam generation under nominal flow conditions for a
generic fission power distribution. The in-core mass flow distribution
results for steady state nominal conditions are in agreement with the
expected from design, thus getting a first assessment of the coupled
1/3D model. Results for nominal condition were compared with
those obtained with a previous 1/3D single-phase model getting more
realistic temperature patterns, also allowing visualize low values of
void fraction inside the upper plenum. It must be mentioned that the
current results were obtained by imposing prescribed fission power
functions from literature. Therefore, results are showed with the aim
of point out the potentiality of the developed model.", keywords = "CFD, PHWR, Thermo-hydraulic, Two-phase flow.", volume = "9", number = "11", pages = "659-6", }