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CFD simulation and analysis of reactor integral hydraulic tests
Shuo Chen, Hanyang Gu
⇑
School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240, China
a r t i c l e i n f o
Article history:
Received 3 May 2019Received in revised form 12 June 2019Accepted 27 July 2019
Keywords:
CFDFlow distributionScaling velocityCore resistance distributionLateral resistance
a b s t r a c t
Scaled-down reactor models are widely used to investigate the reactor integral hydraulic characteristics,due to the feasibility and economy. However, no deﬁnite conclusion can be drawn from previousresearches for the effects of different scaling methods on simulation results. In this paper, in order to ana-lyze the effects of simulation core design parameters and ﬂow ﬂux on core inlet ﬂow distribution, a CFDsimulation is conductedbased on thehydraulic testsof the1/4 scale modelof 600 MWreactor inQinshanphase II. The effects of different scaling velocities, core resistance distributions and lateral resistance onﬂow distribution are investigated. Calculation results reveal that the uniformity of ﬂow distributionincreases with test velocity. Ignoring the resistance of fuel assembly would reduce the uniformity of ﬂowdistribution. The ﬂow distributions under conditions of uniformly and segmented distributed core resis-tance are basically the same, indicating that the detailed simulation of axial pressure drop distribution inassemblies is unnecessary for the design of hydraulic simulator of assembly. The lateral resistance couldsigniﬁcantly affect the ﬂow distribution and excessive lateral resistance would reduce the uniformity of distribution.
2019 Elsevier Ltd. All rights reserved.
1. Introduction
Thermal hydraulic characteristics of reactor core are directlyaffected by the ﬂow distribution at the core inlet, which wouldinﬂuence the operating limits of nuclear power plants ( Jeonget al., 2005). Thus, veriﬁcation tests of core inlet ﬂow distributionmust be carried out in the process of safety review for the designof core of new-structure reactor. In consideration of the feasibilityand economy of tests, the scaled-down reactor models are widelyused to investigate the reactor integral hydraulic characteristics.Takingvariousinﬂuencingfactorsinto account,differentscalingmethods were adopted in different research institutions aroundthe world. In terms of the scaling velocity, linear scaling methodol-ogywasusedinthe1/4scalemodelofQinshanphaseIIreactorandthe scaling velocity was 0.6 times of that of the prototype (Yanget al., 2003). Modiﬁed linear scaling methodology was applied inthe 1/5 scale model of APR + reactor and
ﬃﬃﬃ
5
p
=
5 times of prototypevelocity was selected as the scaling velocity (Euh et al., 2011;Euh et al., 2012a). Unlike two scaling methods mentioned above,researchers kept the scaling velocity the same as that of the proto-type in the 1/5 scale model of SMART reactor (Euh et al., 2012b).For the simulation of reactor core, there were different ways todeal with the core resistance according to previous studies. In Qin-shan phase-II reactor model, the core was divided into several seg-ments, of which the resistance was the same as that of theprototype respectively (Wang and Zong, 2000). Remarkably, theresistances were different for each segment. In APR + reactormodel, the core pressure drop was scaled with some scaling lawsand four perforated plates were regularly installed in the core sim-ulator to have a uniform resistance distribution (Bae et al., 2011).However, the core regions were empty in US-APWR reactor model.The resistance of fuel assembly was ignored in the tests(Watanabe, 2008).In present literatures, no deﬁnite conclusion is made for theeffects of different scaling methods of velocity and core resistanceon ﬂow distribution. In this paper, the 1/4 scale model of 600 MWreactor in Qinshan phase II is taken for research objects. The effectsof scaling velocity, resistance distribution and lateral resistance onthe core inlet ﬂow distribution are numerically investigated bycomputational ﬂuid dynamics (CFD). The simulation results arevalidated with test data. The analysis can provide reference forthe reactor integral hydraulic tests.
2. Numerical approach
2.1. Geometry
Fig. 1 shows the geometry of calculation regions, including twoinlets, downcomer, lower plenum, core inlet dome, core, upper
https://doi.org/10.1016/j.anucene.2019.1069620306-4549/
2019 Elsevier Ltd. All rights reserved.
⇑
Corresponding author.
E-mail address:
guhanyang@sjtu.edu.cn (H. Gu).Annals of Nuclear Energy 135 (2020) 106962
Contents lists available at ScienceDirect
Annals of Nuclear Energy
journal homepage: www.elsevier.com/locate/anucene
plenum and two outlets. Since the test geometry is relatively com-plex, some simpliﬁcation is utilized in several parts of calculationregions. The heat shielding boards in small scale and thin thicknessare simpliﬁed, due to the little effect on the ﬂow distribution(Zhang et al., 2010a,b). Before the simpliﬁcation, 484 holes arelocated in the core inlet dome and every four of them arecorresponding to each of the 121 fuel assemblies in the core.According to the area equivalent principle, 121 simpliﬁed holes
Fig. 1.
Schematic of calculation regions.
0 2 4 6 8 10 12 143.73.83.94.04.14.24.34.44.54.64.74.8
Column G Mesh-A Mesh-B Mesh-C
M a s s F l o w R a t e [ k g / s ]
Assembly Number
Fig. 2.
Mesh independence.
Table 1
Calculation settings of all cases.
Case U
inlet
(m/s)U
inlet
/U
p
MinimumReynoldsnumberCoreResistanceDistributionC
R2l
=
C
R2a
1 4.35 1/4 47,960 Uniform Distribution 102 8.70 1
=
ﬃﬃﬃ
4
p
95,920 Uniform Distribution 103 10.44 0.6 115,100 Uniform Distribution 104 17.40 1 191,830 Uniform Distribution 105 10.44 0.6 115,100 No Resistance —6 10.44 0.6 115,100 Segmented Distribution 107 10.44 0.6 115,100 Uniform Distribution 28 10.44 0.6 115,100 Uniform Distribution 20
Nomenclature
C
R1
linear resistance coefﬁcient, kg/(m
3
s)
C
R2
quadratic resistance coefﬁcient, kg/m
4
K
loss
empirical loss coefﬁcient, 1/m
K
perm
permeability, m
2
n
number of assemblies
q
i
normalized mass ﬂux of assembly
iQ
i
mass ﬂux of assembly
i
, kg/s
U
coreinlet
core inlet velocity, m/s
U
inlet
inlet velocity, m/s
U
p
prototype velocity, m/s
V
physical volume of the porous medium, m
3
V’
volume available to ﬂow, m
3
D
p
core
core pressure drop, Pa
Greek symbols
c
volume porosity
l
dynamic viscosity, Pa
s
q
density, kg/m
3
f
local resistance coefﬁcient, 1/m
Subscripts
a axiall lateral
Fig. 3.
Adopted mesh of the reactor.2
S. Chen, H. Gu/Annals of Nuclear Energy 135 (2020) 106962
Fig. 4.
Streamlines in reactor model.
Fig. 5.
Calculation results of normalized mass ﬂux.
S. Chen, H. Gu/Annals of Nuclear Energy 135 (2020) 106962
3
are substituted for the 484 holes to reduce meshing difﬁculties andcomputing time. What’s more, porous media is introduced to sim-ulate the complicated core region which is ﬁlled with lots of fuelassemblies. The upper plenum imposes little inﬂuence on the ﬂowdistribution, due to the great distancesfrom thecore inlet. Thus, anempty cylinder is selected to modify the upper plenum. Unlike themodiﬁed regions mentioned above, the lower plenum composedwith combined supporting frames and some other components isnot simpliﬁed resulting from their signiﬁcant effects on the ﬂow.In the whole geometry structure, only the core region presents agood symmetry.
2.2. Mesh generation and independency
The mesh generations in all computational regions are dis-played in Fig. 2. Combining the structured and the unstructuredgrids, different types of mesh are utilized in different calculationregions. Due to the irregularity of shape or complexity in structure,the tetrahedral mesh is utilized in the lower plenum and ﬂowregions near the inlet and outlet. In other parts of regular shape,the hexahedral mesh is utilized for reducing the calculationresource. As reported by Yan et al. (2012), General Grid Interface(GGI) could provide satisﬁed connection between different typesof mesh elements and allow non-uniformity of node location onthe interface. Hence, the GGI connection is adopted in the connec-tion of various mesh types in the meshing progress. The core inletﬂow distribution is greatly inﬂuenced by the severe mixing of ﬂowﬂuidin thelowerplenum.As a resultof connectingthelowerﬂuidswith the core, supporting columns also have a signiﬁcant effect onﬂow distribution. Therefore, mesh reﬁnements are utilized in theseregions.To guarantee the independence of the calculated results withthe mesh numbers, three kinds of mesh are utilized in the sensitiv-ity study. The total numbers of mesh elements are 12 million(Mesh-A), 34 million (Mesh-B) and 65 million (Mesh-C) respec-tively. Calculations are performed under the same conditions givenincase 3 in Table 1. As illustrated inFig. 2, the calculatedmassﬂow
rateinColumnG (Thelocationis giveninFig.5)isselectedto makethe comparison. It is found that the difference between results of Mesh-B and Mesh-C is considerably less. However, the simulationresult of Mesh-A is not in good consistent with the others shownin
Fig. 6.
Deviations between calculation and test values of normalized mass ﬂux.
-16-14-12-10-8-6-4-20246810121416051015202530354045
F r e q u e n c y
% Difference (Calculation to Test)
Fig. 7.
Statistic results of calculated deviations.4
S. Chen, H. Gu/Annals of Nuclear Energy 135 (2020) 106962

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