Research Papers

Computational Fluid Dynamic Simulation of the Moderator Flow in CANDU-6 Nuclear Reactors

[+] Author and Article Information
Foad Mehdi Zadeh

Department of Engineering Physics,
Polytechnique Montréal,
Montréal, Québec H3T 1J4, Canada
e-mail: foad.mehdi-zadeh@polymtl.ca

Stéphane Etienne

Department of Mechanical Engineering,
Polytechnique Montréal,
Montréal, Québec H3T 1J4, Canada
e-mail: stephane.etienne@polymtl.ca

Alberto Teyssedou

Department of Engineering Physics,
Polytechnique Montréal,
Montréal, Québec H3T 1J4, Canada
e-mail: alberto.teyssedou@polymtl.ca

1Corresponding author.

Manuscript received October 9, 2015; final manuscript received February 18, 2016; published online December 20, 2016. Assoc. Editor: Andrey Churkin.

ASME J of Nuclear Rad Sci 3(1), 011010 (Dec 20, 2016) (13 pages) Paper No: NERS-15-1207; doi: 10.1115/1.4032874 History: Received October 09, 2015; Accepted February 21, 2016

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.

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Grahic Jump Location
Fig. 5

Experimental setup and positions of the measurement planes given in Paul [35] and Paul et al. [36]

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Fig. 6

Comparison of simulated lateral velocity profile with data of Paul et al. [36,38]: (a) x/d=1.25, (b) x/d=3.35, (c) x/d=5.45, and (d) x/d=7.55

Grahic Jump Location
Fig. 4

Typical averaged channel thermal power [31]

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Fig. 3

Grid topology around (a) Calandria tubes and (b) water injectors

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Fig. 2

Cross-sectional view of the Calandria vessel of a CANDU-6 nuclear power reactor

Grahic Jump Location
Fig. 1

Validity of the Boussinesq approximation at T0=71°C (nominal CANDU-6 operating value); where ϵ9 is calculated by Eq. (5), the coefficient 1044 has units of m/°C. The temperature difference is calculated with respect to the outlet moderator condition.

Grahic Jump Location
Fig. 7

Expected moderator flow configurations: (a) momentum-dominated, (b) mixed-type, and (c) buoyancy-dominated

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Fig. 8

Sampled values for the inertia-dominated configuration (Ri=0.008): 1, y=+1.43  m; 2, y=0  m; 3, y=−1.43  m; 4, x=+1.43  m; 5, x=0  m; and 6, x=−1.43  m

Grahic Jump Location
Fig. 9

Sampled values for mixed-type configuration (Ri=0.07): 1, y=+1.43  m; 2, y=0  m; 3, y=−1.43  m; 4, x=+1.43  m; 5, x=0  m; and 6, x=−1.43  m

Grahic Jump Location
Fig. 10

Sampled values for buoyancy-dominated configuration (Ri=0.23): 1, y=+1.43  m; 2, y=0  m; 3, y=−1.43  m; 4, x=+1.43  m; 5, x=0  m; and 6, x=−1.43  m

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Fig. 11

Qualitative comparison of velocity vectors for Ri=0.05: (a) predicted by Code_Saturne and (b) predicted by MODTURC [7]

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Fig. 12

Moderator flow configuration map. (a) Proposed by Carlucci and Cheung [7] and (b) present work using Code_Saturne.




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