For CANada Deuterium Uranium (CANDU) nuclear reactors, the characterization of the moderator thermal-hydraulic behavior under both normal and abnormal operating conditions constitutes an important safety issue. For normal operating conditions, the flow temperature distribution may produce changes on the heavy-water mass density, which in turn may affect the neutron moderation rate. Consequently, these variations influence the thermal neutron flux distribution in the reactor core. Therefore, it is fundamental to know all possible moderator flow configurations as well as the corresponding temperature distributions. In particular, any possibility of a dryout at the external wall of the Calandria tubes and consequently excessive temperature excursions must be prevented. Within this framework, this paper presents detailed two-dimensional (2D) numerical steady-state simulations for a wide range of flow conditions. Both the accuracy of the numerical approximations and the validity of some physical models used in computational fluid dynamic (CFD) codes are assessed. The numerical results are then used to construct a cartographical representation of moderator flows in CANDU-6 reactors. To support the existence of coherent flow asymmetries and eventually flow-structure oscillations, the present numerical results are also compared with the previous ones obtained using a porous medium-modeling approach.

References

References
1.
IAEA
, “
Power Reactor Information System
,” http://www.iaea.org/PRIS/WorldStatistics/OperationalReactorsByCountry.aspx (Accessed: Apr. 1, 2014).
2.
Cacuci
,
D. G.
,
2010
,
Handbook of Nuclear Engineering
,
Springer
.
3.
Carlucci
,
L.
,
1982
, “
Numerical Simulation of Moderator Flow and Temperature Distributions in a CANDU Reactor Vessel
,”
Proceedings of La Modélisation fine des Écoulements
,
Presses de l'école nationale des ponts et chaussées
,
Paris
, pp. 
533
543
, ISBN: 2829780467.
4.
Gosman
,
A.
, and
Pun
,
W.
,
1974
, “
Calculation of Recirculating Flows
,”
Imperial College
,
London
, .
5.
Mandal
,
J. C.
, and
Sonawane
,
C. R.
,
2014
, “
Simulation of Moderator Flow and Temperature Inside Calandria of CANDU Reactor Using Artificial Compressibility Method
,”
Heat Transfer Eng.
,
35
(
14–15
), pp. 
1254
1266
. 0145-763210.1080/01457632.2013.876802
6.
Szymanski
,
J.
,
Garceau
,
M.
,
Ng
,
K.
, and
Midvidy
,
W.
,
1983
, “
Numerical Modelling of Three-Dimensional Turbulent Moderator Flow in Calandria
,”
Proceedings of Numerical Methods in Nuclear Engineering
, Vol. 
1
,
Canadian Nuclear Society
,
Montréal
, pp. 
970
84
.
7.
Carlucci
,
L.
, and
Cheung
,
I.
,
1986
, “
The Effects of Symmetric/Asymmetric Boundary Conditions on the Flow of an Internally Heated Fluid
,”
Numer. Methods Part. Diff. Equat.
,
2
(
1
), pp. 
47
61
.10.1002/(ISSN)1098-2426
8.
Huget
,
R.
,
Szymanski
,
J.
, and
Midvidy
,
W.
,
1989
, “
Status of Physical and Numerical Modelling of CANDU Moderator Circulation
,”
Proceedings of 10th Annual Conference of the Canadian Nuclear Society
,
Canadian Nuclear Society
,
Ottawa, ON
.
9.
Huget
,
R.
,
Szymanski
,
J.
,
Galpin
,
P.
, and
Midvidy
,
W.
,
1990
. “
Modturc_clas: An Efficient Code for Analyses of Moderator Circulation in CANDU Reactors
,”
Proceedings of 3rd International Conference on Simulation Methods in Nuclear Engineering
,
Canadian Nuclear Society
,
Montreal
.
10.
Carlucci
,
L.
,
Agranat
,
V.
,
Waddington
,
G.
,
Khartabil
,
H.
, and
Zhang
,
J.
,
2000
. “
Predicted and Measured Flow and Temperature Distribution in a Facility for Simulating In-Reactor Moderator Circulation
,”
Proceedings of 8th International Conference of CFD Canada
,
CFD Society of Canada
,
Montreal
.
11.
Yoon
,
C.
,
Rhee
,
B. W.
, and
Min
,
B.-J.
,
2004
, “
3D CFD Analysis of the CANDU-6 Moderator Circulation Under Normal Operating Conditions
,”
J. Korean Nucl. Soc
,
36
(
6
), pp. 
559
570
. 0372-7327
12.
Yoon
,
C.
,
Rhee
,
B. W.
, and
Min
,
B.-J.
,
2004
, “
Development and Validation of the 3-D Computational Fluid Dynamics Model for CANDU-6 Moderator Temperature Predictions
,”
Nucl. Technol.
,
148
(
3
), pp.
259
267
. 0029-5450
13.
Yoon
,
C.
, and
Park
,
J. H.
,
2008
, “
Development of a CFD Model for the CANDU-6 Moderator Analysis Using a Coupled Solver
,”
Ann. Nucl. Energy
,
35
(
6
), pp. 
1041
1049
.10.1016/j.anucene.2007.11.013
14.
Arsene
,
R.
,
Prisecaru
,
I.
, and
Nicolici
,
Ş.
,
2013
, “
Improvement of the Thermalhydraulic Characteristics in the Calandria Vessel of a CANDU 6 Nuclear Reactor
,”
UPB Sci. Bull.
,
75
(
4
), pp.
251
262
.
15.
Rhee
,
B. W.
, and
Kim
,
H. T.
,
2014
. “
A Review of the Scaling Study of the CANDU-6 Moderator Circulation Test Facility
,”
J. Power Energy Eng.
,
2
(
09
), pp. 
64
73
.10.4236/jpee.2014.29010
16.
Kim
,
M.
,
Yu
,
S.-O.
, and
Kim
,
H.-J.
,
2006
. “
Analyses on Fluid Flow and Heat Transfer Inside Calandria Vessel of CANDU-6 Using CFD
,”
Nucl. Eng. Des.
,
236
(
11
), pp. 
1155
1164
.10.1016/j.nucengdes.2005.10.018
17.
Sarchami
,
A.
,
Ashgriz
,
N.
, and
Kwee
,
M.
,
2014
, “
Three Dimensional Numerical Simulation of a Full Scale CANDU Reactor Moderator to Study Temperature Fluctuations
,”
Nucl. Eng. Des.
,
266
, pp. 
148
154
.10.1016/j.nucengdes.2013.11.042
18.
Farhadi
,
F.
,
Ashgriz
,
N.
,
Kwee
,
M.
,
Girard
,
R.
,
Parlatan
,
Y.
, and
Ali
,
M.
,
2009
. “
Temperature Fluctuations in a CANDU Moderator Test Facility
,”
Proceeding of International Conference on Nuclear Engineering
,
ASME
,
Brussels
, pp. 
569
577
.
19.
Kim
,
H. T.
, and
Rhee
,
B. W.
,
2015
, “
Scaled-Down Moderator Circulation Test Facility at Korea Atomic Energy Research Institute
,”
Sci. Technol. Nucl. Installations
, pp.
1
10
, Article ID 760870, in press.
20.
Chorin
,
A. J.
,
1967
, “
A Numerical Method for Solving Incompressible Viscous Flow Problems
,”
J. Comput. Phys.
,
2
(
1
), pp. 
12
26
.10.1016/0021-9991(67)90037-X
21.
Fluid Dynamics, Power Generation and Environment Department, Single Phase Thermal-Hydraulics Group
,
Code_Saturne 3.0.0 Theory Guide
, http://code-saturne.org/cms/sites/default/files/theory-3.0.pdf (Accessed: Feb. 10, 2014).
22.
Boussinesq
,
J.
,
1901
,
Théorie analytique de la chaleur: mise en harmonie avec la thermodynamique et avec la théorie mécanique de la lumière
, Vol. 
1
,
Gauthier-Villars
,
Paris, France
.
23.
Gray
,
D. D.
, and
Giorgini
,
A.
,
1976
. “
The Validity of the Boussinesq Approximation for Liquids and Gases
,”
Int. J. Heat Mass Transfer
,
19
(
5
), pp. 
545
551
.10.1016/0017-9310(76)90168-X
24.
Nero
,
A. V.
,
1979
.
A Guidebook to Nuclear Reactors
,
University of California Press
,
Berkeley and Los Angeles, CA
.
25.
Archambeau
,
F.
,
Méchitoua
,
N.
, and
Sakiz
,
M.
,
2004
. “
Code Saturne: A Finite Volume Code for the Computation of Turbulent Incompressible Flows-Industrial Applications
,”
Int. J. Finite Vol.
,
1
(
1
), pp.
1
62
.
26.
Favre
,
A.
,
1976
,
La turbulence en mécanique des fluids
,
Gauthier-Villars
,
Paris, France
.
27.
Teyssedou
,
A.
,
Necciari
,
R.
,
Reggio
,
M.
,
Mehdi Zadeh
,
F.
, and
Etienne
,
S.
,
2014
. “
Moderator Flow Simulation Around Calandria Tubes of CANDU-6 Nuclear Reactor
,”
Eng. Appl. Comput. Fluid Mech.
,
8
(
1
), pp. 
178
192
.
28.
Launder
,
B. E.
, and
Spalding
,
D.
,
1974
, “
The Numerical Computation of Turbulent Flows
,”
Comput. Methods Appl. Mech. Eng.
,
3
(
2
), pp. 
269
289
.10.1016/0045-7825(74)90029-2
29.
Van Doormaal
,
J.
, and
Raithby
,
G.
,
1984
, “
Enhancements of the Simple Method for Predicting Incompressible Fluid Flows
,”
Numer. Heat Transfer
,
7
(
2
), pp. 
147
163
.10.1080/01495728408961817
30.
Bouquillon
,
M.
,
2008
, “
Modélisation numérique de jets et leurs applications dans la simulation des écoulements dans la cuve du modérateur du réacteur CANDU
,” M.Sc. Thesis,
Polytechnique Montréal
.
31.
Rozon
,
D.
,
2007
,
Gestion du combustible nucléaire
,
Polytechnique Montréal
, .
32.
Sarchami
,
A.
,
2011
, “
Investigation of Thermal Hydraulics of a Nuclear Reactor Moderator
,” Ph.D. Thesis,
University of Toronto
.
33.
Richardson
,
L. F.
,
1911
, “The Approximate Arithmetical Solution by Finite Differences of Physical Problems Involving Differential Equations, With an Application to the Stresses in a Masonry Dam,”
Philosophical Transactions of the Royal Society of London. Series A: Containing Papers of a Mathematical or Physical Character
,
The Royal Society Publishing
, pp. 
307
357
.
34.
Celik
,
I.
,
2004
, “
Procedure for Estimation and Reporting of Discretization Error in CFD Aaplications
,”
ASME J. Fluids Eng.
,
1
(
06
), pp. 
2008
.
35.
Paul
,
S. S.
,
2007
, “
Experimental and Numerical Studies of Turbulent Cross-Flow in a Staggered Tube Bundle
,” M.Sc. Thesis,
University of Manitoba
.
36.
Paul
,
S.
,
Tachie
,
M.
, and
Ormiston
,
S.
,
2007
, “
Experimental Study of Turbulent Cross-Flow in a Staggered Tube Bundle Using Particle Image Velocimetry
,”
Int. J. Heat Fluid Flow
,
28
(
3
), pp. 
441
453
.10.1016/j.ijheatfluidflow.2006.06.001
37.
Ansys
,
2009
,
Ansys Fluent 12.0 Users Guide
,
Ansys Inc
,
Canonsbug, PA
.
38.
Paul
,
S.
,
Ormiston
,
S.
, and
Tachie
,
M.
,
2008
, “
Experimental and Numerical Investigation of Turbulent Cross-Flow in a Staggered Tube Bundle
,”
Int. J. Heat Fluid Flow
,
29
(
2
), pp. 
387
414
.10.1016/j.ijheatfluidflow.2007.10.001
39.
Manzer
,
A.
,
1979
, “
Design Manual Gentilly-2 Nuclear Generating Station
,”
Atomic Energy of Canada Limited
, .
You do not currently have access to this content.