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Nuclear Engineering and Design 229 (2004) 2546
Analysis of natural circulation phenomena in VVER-1000
Seyed Khalil Mousavian a,b,, Francesco DAuria b, Mahmoud A. Salehi a
a Department of Mechanical Engineering, Sharif University of Technology, NPP of AEOI, P.O. Box 11365-9567 Tehran, Iranb Dipartimento di Ingegneria Meccanica Nucleare e Della Produzione, Universit di Pisa, Via Diotisalvi 2, I-56126 Pisa, Italy
Received 1 July 2003; received in revised form 13 October 2003; accepted 18 November 2003
Abstract
In all light water reactors (LWR), natural circulation is an important passive heat removal mechanism. In the present pa-
per, the natural circulation phenomena are studied with reference to step-wise coolant inventory reduction and a small break
loss-of-coolant-accident (SBLOCA) in the cold leg of VVER-1000. The natural circulation flow map (NCFM) approach is con-
sidered to evaluate the natural circulation performance of the VVER-1000 NPP also comparing VVER-1000 and PWR systems.
Three different elevations between heat source (core) and heat sink (steam generators) zones have been considered in order to
characterize the buoyancy force in a VVER-1000. The influence of power and the cold legs loop seal upon the natural circulation
performance is also evaluated. In the second part, a series of SBLOCA simulations with break area ranging from 0.5 to 11.7% of
the cold leg cross sectional area are performed starting with the VVER-1000 system in nominal conditions. The effect of Emer-
gency Core Cooling System (ECCS) including passive and active parts of ECCS are evaluated. The simulations were performed
by the help of the system code RELAP5. Within the framework of the qualification of the adopted computational tools, the
results are compared with experimental data from Kozloduy NPP unit 6 test and PSB-VVER integral test facility available fromthe literature. Namely, the qualification of the adopted nodalisation in steady state conditions is achieved by using experimental
data. The accuracy of selected results have been estimated in quantitative terms by applying the fast Fourier transform based
method (FFTBM). Finally, the relevance and the potential for the occurrence of the reflux condensation mode, i.e., one of the
Natural Circulation regimes, for cooling of reactor core in VVER-1000 are discussed.
2003 Elsevier B.V. All rights reserved.
1. Introduction
Natural circulation phenomena play an important
role in energy extraction from hot zones and rejectionto the cold zones without using a mechanical pump
in several engineering systems, such as solar water
heaters, geothermal heat extraction, electrical machine
motor cooling and reactor core cooling (Zvirin, 1981).
The density difference between fluid leaving the heat
Corresponding author. Tel.: +39-050-836675;
fax: +39-050-836665.
E-mail address: kh [email protected]
(S.K. Mousavian).
source at one elevation and the heat sink(s) in higher
elevation induces flow and provides a passive means
of core cooling.
Naturally driven systems have been used in theexisting nuclear reactors both to remove the core
heat under normal operation conditions including
full power operation (like the Russian boiling wa-
ter reactor VK-50 or the Dodewaard reactor in the
Netherlands) and to fulfil some safety functions (Hart
et al., 2001; Warnke et al., 2001). Many experiments
have been performed to study of the natural circula-
tion phenomena in Western pressurized water reactor
(PWR) test facilities (DAuria and Galassi, 1992;
Loomis and Soda, 1982; Tasaka et al., 1988; Ferng
0029-5493/$ see front matter 2003 Elsevier B.V. All rights reserved.
doi:10.1016/j.nucengdes.2003.11.012
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26 S.K. Mousavian et al. / Nuclear Engineering and Design 229 (2004) 2546
Nomenclature
A flow area
AA average amplitudeACCU accumulators
AS wall cross sectional area
ECCS emergency core cooling system
F function
f frequency in FFT analysis
FFT fast Fourier transform
FW feed waterG mass fluxg gravity acceleration
HPIS high pressure injection system
ISB integral test facilityL length
l axial length
LPIS low pressure injection system
MCP main coolant pump
MSH main steam headerN number of points in FFT analysis
NCFM natural circulation flow map
NPP nuclear power plant
NC natural circulation
PACTEL integral test facility
PWR pressurized water reactor
PMK integral test facilityPSB integral test facility
RELAP reactor excursion and leak analysis
program
REWET integral test facility
RPV reactor pressure vessel
RM residual mass
SBLOCA small break loss-f-coolant-
accident
SG steam generator
St. St. steady state
t timeVVER water-cooled water-moderated
reactor
w weighting factor
WF weighted frequency
Z elevation
Greek letters
difference symbolsummation symbol
F difference between experiment
and calculation function in
frequency domain
P pressure loss
Z average elevation difference
friction coefficient
Subscripts
accl acceleration term
B buoyancy
cal calculation
CL cold leg
CL-HL between CL and HL
exp experimental
f frictional termHL hot leg
j index of summation in FFT analysis
k index of summation through loop
m power of 2 in FFT analysis
mo model
mix mixture in two-phase flown index of summation in FFT analysis
norm normalization
pipelines pipe lines through loop
pr proto-type
pump-off pump is offRPV reactor pressure vessel
R ratio
saf safety
SGs steam generators
var index of number of parameters in
FFT analysis
and Lee, 1994). Other studies focused on evaluation
and modeling of natural circulation behavior dur-
ing small break loss-of-coolant-accident (SBLOCA)
and transients in Western-type reactors (Duffey andSursock, 1987; Hsu et al., 1990; Mousavian and
Salehi, 2001).
The natural circulation experimental database in
VVER geometry, especially in VVER-1000, is lim-
ited if compared with the database available for
Western PWRs. In VVER-440 geometry experiments
were performed in REWET-III, PMK and PACTEL
integral test facilities (Kervinen and Hongisto, 1986;
Toth, 1988; Lomperski and Kouhia, 1994; Puustinen,
2002). In the case of VVER-1000 there are two inte-
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S.K. Mousavian et al. / Nuclear Engineering and Design 229 (2004) 2546 27
gral test facilities: ISB (small scale, with a two-loop
model of primary circuit) and PSB (large scale, with a
four-loop model of primary circuit), both of those in-
stalled at Elektrogorsk in Russia. Natural circulationexperiments relevant to SBLOCA and the stepwise
reduction of mass inventory are reported for ISB
(Krepper, 1997, 1999). So far, some data related to
natural circulation in PSB-VVER integral test facility
were published (Blinkov et al., 2003). In addition, the
VVER-1000 natural circulation database is enlarged
by the results from the Kozloduy NPP unit 6 program
(Pavlova et al., 2000).
In Western type PWRs by decreasing mass inven-
tory three or four main modes of natural circulation
are identified: single phase flow, stable two phase
flow, unstable two phase flow with siphon effect andboiler-condenser or reflux condensation. The primary
circuit mass flow rate is a strong function of buoy-
ancy pressure head, pressure losses, mass inventory
or void fraction, and a weak function of power.
Two main objectives of this paper can be distin-
guished:
Natural circulation performance study in VVER-
1000; In depth study of natural circulation phenomena in
complex systems.
Related to the second objective, reflux condensa-
tion cooling is significant at small mass inventories
of primary system (e.g., Kawanishi et al., 1991). Dur-
ing SBLOCA, reflux condensation does not contribute
much to heat removal when the break is larger than
a defined value, around 1% of cold leg size in West-
ern PWR, as reported by Jeong (2002). The loop seal
effect on heat removal capability and qualification of
natural circulation conditions is noticeable (DAuria
and Frogheri, 2002; Krepper, 1999).
In order to achieve both the defined objectives, qual-ification of the adopted tools and of the analysis pro-
cedure is needed.
2. Nodalization and qualification
The Relap5/Mod3 code is used for the present anal-
yses to simulate natural circulation phenomena in a
VVER-1000. This code is based on one-dimensional
six-equation models and was developed for best-
estimate simulation of thermalhydraulic behavior of
LWRs during accidents and transients. The adopted
nodalization of one loop of primary and secondary
sides is shown in Fig. 1.The nodalisation qualification is achieved by a com-
plex process as discussed by DAuria and Galassi
(1998), where the steady-state and the on-transient
qualifications levels are distinguished. A sample result
derived from the steady-state nodalisation qualifica-
tion process is given in Fig. 2. Selected steps and re-
sults from the on-transient nodalisation qualification
are discussed below (Sections 2.1 and 2.2).
2.1. Comparison and quantitative assessment with
Kozloduy NPP unit 6
Natural circulation test data from Kozloduy NPP
unit 6 were measured by Bulgarian and Russian ex-
perts during the plant commissioning phase (Pavlova
et al., 2000). Initial and boundary conditions related to
one of the selected test conditions are given in Table 1.
Fig. 3 shows the comparison between results from
the VVER-1000 code-nodalisation and those from Ko-
zloduy NPP unit 6 test. Two code calculations have
been performed, owing to inadequate knowledge of
all boundary conditions of the NPP test: in case one,
main feed waters (FW) flow rates is 20% of nominalvalue and initial pressurizer level is 7.2 m; in case two,
mass flow rates of main FW is 10% of nominal value
and initial pressurizer level is 5.2 m.
From a qualitative point of view, the calculated re-
sults are judged to be sufficiently close to the measured
data, also considering the unknown uncertainty that
characterizes the measurements. Therefore, it was de-
cided to apply the fast Fourier transform based method
(FFTBM; Ambrosini et al., 1990; DAuria et al., 2000;
Prosek et al., 2002) in order to quantify the accuracy,
even though the FFTBM tool is normally applied whenan experimental database wider that the present one is
available.
The accuracy quantification of a code calculation
considers the amplitude, in the frequency domain, of
the experimental signal Fexp(t) and the error function:
F Fcal(t) Fexp(t) (1)
In particular, the method defines two values which
are characteristic of every calculation: average ampli-
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28 S.K. Mousavian et al. / Nuclear Engineering and Design 229 (2004) 2546
Fig. 1. RELAP5/mod3.2 nodalization for NC study in a VVER-1000.
tude (AA) and weighted frequency (WF);
AA =
N/2=2mn=0 |F(fn)|N/2=2m
n=0 |Fexp(fn)|and
WF =
N/2=2mn=0 |F(fn)|fnN/2=2m
n=0 |F(fn)|(2)
The overall accuracy is obtained by defining average
performance indexes, total average amplitude:
AAtot =Nvarj=1
(AA)j (wf)j, (WF)tot
=
Nvarj=1
(WF)j (wf)j, with
Nvarj=1
(wf)j = 1
(3)
(wf)j =(wexp)j (wsaf)j (wnorm)jNvar
j=1 (wexp)j (wsaf)j (wnorm)j(4)
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S.K. Mousavian et al. / Nuclear Engineering and Design 229 (2004) 2546 29
Table 1
Initial and boundary conditions and sequence of events at Kozloduy unit 6 test
Initial and boundary conditions Sequence of events
Parameters Unit Values Time (s) Values/conditions
Reactor power MW 151 0 All MCPs are switched off
Primary side pressure MPa 15.70
Pressure in MSH MPa 6.13 60 Auxiliary FW pumps start
Pump heads MPa 0.62
Pressurizer water level m 5.20 90 Control rod group #8 starts withdrawal
Hot/cold legs temperatures C 283.5/281
PRZ steam temperature C 344.2 220 Auxiliary FW flow rate reaches to 21/42 kg/s
SG water level m 2.45
Temperature of main FW C 161.7 385 Make-up flow rate reaches to 9 kg/s
Make up/let-down flow rates kg/s 8.4/8.4
Emergency FW flow rates kg/s 44.5/36.5 590 Let-down flow rate reaches to 5.5 kg/s
SG pressure MPa 6.13
Core exit temperature C 282.1 630 End of experiment
where Nvar is the number of the parameters analyzed,
and (AA)j, (WF)j and (wf)j are average amplitude,
weighted frequency and weighting factors for i-th ana-
lyzed parameter, respectively. Each (wf)j accounts for
experimental accuracy, safety relevance of particular
parameters and its relevance with respect to pressure
(or primary pressure normalization). Weighting factor
components of Eq. (4) are shown in Table 2 (Bovalini
et al., 1992).
For FFTBM calculations the primary side pressure,hot and cold leg temperatures, pressurizer water level,
and steam generator down-comer level are selected.
Table 3 shows the results of calculations for the two
-20
0
20
40
60
80
100
120
140
-0.5 0 0.5 1 1.5 2 2.5 3 3.5 4 4.5
Height (m)
Volume(m^3)
RELAP5
EXPERIMENT
Fig. 2. Qualification of nodalization at steady state level for sec-
ondary side of SG.
cases as mentioned. We assumed N/2 = 212, also cut
frequencies (fcat) 1, 3 and 5 Hz are selected in FFTBM
calculations.
With reference to the accuracy of a given calcula-
tion, the following acceptability criterion is defined:
AAtot K (5)
where K is acceptability factor valid for whole the
transient. Based on several calculations K = 0.4 hasbeen fixed (DAuria et al., 1996). It can be noted:
1. AAtot 0.3 characterize very good predictions,
2. 0.3 AAtot 0.5 characterize good code predic-
tions,
3. 0.5 AAtot 0.7 characterize poor code predic-
tions,
4. AAtot 0.7 characterize very poor code predic-
tions.
Table 2
Weighting factor components for the analyzed quantities (Bovaliniet al., 1992)
Quantity wexp wsaf wnorm
Pressure drops 0.7 0.7 0.5
Mass inventories 0.8 0.9 0.9
Flow rates 0.5 0.8 0.5
Primary pressure 1.0 1.0 1.0
Secondary pressure 1.0 0.6 1.1
Fluid temperatures 0.8 0.8 2.4
Clad temperatures 0.9 1.0 1.2
Collapsed levels 0.8 0.9 0.6
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30 S.K. Mousavian et al. / Nuclear Engineering and Design 229 (2004) 2546
15
15.5
16
16.5
17
17.5
18
18.5
19
0 100 200 300 400 500 600Time (s)
Primarypressure(MPa)
case1
case2
Kozloduy
2.32
2.34
2.36
2.38
2.4
2.42
2.44
2.46
2.48
2.5
2.52
0 100 200 300 400 500 600
Time (s)
SGlevel(m)
case1
case2
Kozloduy
3
3.5
4
4.5
5
5.5
6
6.5
7
7.5
8
0 100 200 300 400 500 600
Time (s)
Pressurizerlevel(m)
case1
case2
Kozloduy
260
270
280
290
300
310
320
0 100 200 300 400 500 600
Time (s)
Coldandhotlegtemperatures(C)
case1 hot -leg case1 cold- leg
case2 hot -leg case2 cold- leg
Kozloduy hot-leg Kozloduy cold-leg
(a)
(c) (d)
(b)
Fig. 3. Comparison of case 1 and 2 of the present study with Kozloduy NPP unit 6.
In addition, the acceptability thresholdK =
0.1 hasbeen fixed for the primary pressure, because of its
importance in characterizing any transient scenario.
The most interesting information is given obviously
by AAtot, which represents the relative magnitude of
these discrepancies; WF adds a further information
allowing to better identify the character of accuracy.
As it can be seen from Table 3, the primary pres-
sure criterion (i.e., K = 0.1) was not fulfilled but
the total accuracy is well below 0.3 indicating very
good predictions. The reason for discrepancies be-
tween measured and calculated trends is attributed toinadequate knowledge of the boundary conditions, as
already mentioned. As a conclusion, based on this ap-
plication of the FFTBM, the VVER-1000 nodalisation
was considered qualified at the on-transient level.
2.2. Comparison with PSB-VVER integral test facility
The PSB integral test facility is a four loop, full pres-
sure scaled down model of the primary system of the
NPP with VVER-1000 in Russia. The volume-power
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S.K. Mousavian et al. / Nuclear Engineering and Design 229 (2004) 2546 31
Table 3
Results of quantitative analysis for two cases at different cut
frequenciesa
Variables Case 1 Case 2
AAj WFj AAj WFj
Primary pressure 0.20 0.09 0.23 0.10
0.22 0.26 0.25 0.30
0.22 0.31 0.25 0.36
HL temperature 0.04 0.13 0.04 0.13
0.04 0.38 0.04 0.37
0.04 0.44 0.04 0.43
CL temperature 0.04 0.11 0.05 0.11
0.05 0.34 0.06 0.33
0.05 0.39 0.05 0.39
PRZ collapsed level 0.17 0.05 0.24 0.040.18 0.16 0.24 0.13
0.18 0.19 0.24 0.15
SG collapsed level 0.05 0.07 0.04 0.07
0.05 0.22 0.04 0.21
0.05 0.26 0.04 0.24
Total (AAtot, WFtot) 0.08 0.11 0.09 0.10
0.09 0.31 0.10 0.31
0.09 0.36 0.10 0.36
a Upper, middle and lower numbers are related to 1, 3 and 5 Hz
cut frequencies, respectively.
scale is 1/300 while the elevation scale is 1/1. Fourloops provide an adequate simulating of the emer-
gency and transient conditions leading to asymmetri-
cal thermalhydraulic behavior of the loops. A natural
circulation experiment in PSB has been performed
(Blinkov et al., 2003) and the related main initial and
boundary conditions values are reported in Table 4.
In the PSB-VVER some geometrical similar-
ity criteria including axial length, flow area, wall
Table 4
Initial and boundary conditions in PSB-VVER (Blinkov et al., 2003)
Parameters Unit Initial conditions
(forced conditions)
Natural circulation
conditions
Core power MW 0.500 0.500
Core outlet pressure MPa 15.5 15.7
Core volumetric flow rate m3/s 0.0055
Fluid velocity in the core m/s 0.39
SG pressure MPa 6.5 6.5
Core mass flow rate kg/s 54.3
Core inlet/outlet temperatures C 281.5/283.6 277.0/305.2
SG inlet/outlet temperatures C 283.2/282.1 303.2/281.8
Pump inlet/outlet temperatures C 282.0/281.7 281.4/277.3
cross-sectional area (with the exception of pipelines)
and friction coefficient are considered to be the same
for the model and the proto-type:
lR
lmo
lpr
= 1, AR
Amo
Apr
= 1,
ASR
ASmo
ASpr
= 1, R
mo
pr
= 1 (6)
In reality, the model friction factors exceed the sim-
ilar values in the prototype in each section of the
loop. Thus, the overall pressure drop in the prototype
(VVER-1000) at nominal conditions equals 0.61 MPa,
while in the test facility (PSB-VVER) this value is
0.87 MPa.
Also the scaling criteria which have been developedby Ishii and Kataoka such as Richardson, modified
Stanton, Biot, time ratio and heat source numbers are
satisfied (Ishii and Kataoka, 1983).
The on-transient nodalisation qualification of the
VVER-1000 nodalisation has been performed fol-
lowing the requirements foreseen by the so-called
Kv-scaled calculation (DAuria and Galassi, 1998).
Following this procedure, boundary and initial condi-
tions of the NPP nodalisation are fixed utilizing source
data from the Integral Test Facility, ITF (PSB in the
current case) and proper scaling criteria to scale-upfacility related values. At the end, experimental data
from the ITF are compared with calculation results
from the NPP nodalisation (or input deck).
The comparison between the calculated results from
the VVER-1000 nodalisation and the measured data
in the PSB integral test facility for fluid temperatures
distribution through the loop (given at the time of start
of the experiment as a function of the function of a
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32 S.K. Mousavian et al. / Nuclear Engineering and Design 229 (2004) 2546
260
270
280
290
300
310
320
1 2 3 4 5 6 7 8
Mesh points through loop : 1-core outlet, 2-UP, 3-SG inlet,4-SG outlet, 5-MCP inlet, 6-MCP outlet, 7-dow n comer, 8-core inlet
Temperature(C
)
present study PSB-VVER
265
270
275
280
285
290
295
300
305
310
315
0 100 200 300 400 500 600
Time (s)
FluidtemperaturesatRPVinletandoutlet
(C)
PSB-VVER-CL
PSB-VVER-HL
present-study-CL
present-study-HL
0 50.0 100.0 150.0 200.0 250.0 300.0 350.0 400.0
Time (s)
0
1.00
2.00
3.00
4.00
5.00
6.00
Liquidvelocitythroughcore(m/s)
XXX PSB-VVER
X
X
X
X
X
X
XX
XX X X X X X X X X
YYY present-study
Y
Y
YY
Y Y Y Y Y Y Y Y Y Y Y Y Y Y Y
(a) (b)
(c)
Fig. 4. Comparison of present work with PSB-VVER experimental data.
geometrical abscissa along the loop) including core
inlet and outlet that are given as a function of time
and liquid velocity through the core can be found in
Fig. 4. There is a good agreement between the consid-
ered curves (NPP calculated results and ITF measured
data). No quantitative evaluation of accuracy was done
in this case by the application of the FFTBM.
3. Natural circulation scenario
It is assumed that at time 0 all MCPs trip and re-
actor power scram occurs, so transient conditions for
establishing the natural circulation will initiate. After
about ten minutes (at t = 700 s), mass inventory re-
duction starts, caused by fluid draining of primary side
coolant based upon the steps characterized by Fig. 5.
0.0E+00
2.0E+04
4.0E+04
6.0E+04
8.0E+04
1.0E+05
1.2E+05
1.4E+05
1.6E+05
1.8E+05
-1000 1000 3000 5000 7000 9000 11000 13000 15000
Time (s)
Integralofdrainingmas
sflow
rate(kg)
66% of total mass inventory, at nominal conditions
Fig. 5. Integral of mass flow rate draining.
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S.K. Mousavian et al. / Nuclear Engineering and Design 229 (2004) 2546 33
Table 5
Initial and boundary conditions in NC study
Initial conditions Boundary conditions
Parameters Unit Values Parameters/systems Values/conditions
Reactor power MW 3000.0 ECCS (LPIS, HPIS, ACCU) Not activated
Primary/secondary sides pressures MPa 15.7/6.44 Auxiliary/emergency FW Not activated
Total mass of RPV/primary-side kg 83 103/241 103 Mass flow rate of FWs at: (kg/s)
Steady-state conditions; 163.0
From St.St to 700 s; 16.3
From 700 s to end; 4.9
Fluid temperatures at RPV inlet/outlet C 289/318 All MCPs trip at time, s 0
Fluid temperatures at SG inlet/outlet C 318/288 Starting time of draining, s 700
Pressurizer temperature C 345 Reactor scram at time, s 0
Maximum fuel/cladding temperatures C 340/2100 Mass flow rate of draining: (kg/s)
At time 0 to 800 s 0
From 800 to 3100 s According to Fig. 5
From 3100 s to end 6.0Collapsed level of pressurizer m 8.4
Safety relief valves set points:
Collapsed level of core m 3.25 Opening (MPa) P(SG) 8.24
Closing (MPa) P (SG) 6.86
Collapsed level of SG m 1.8
The 700 s period is needed in order to ensure that MCP
coast down effects are negligible and that the loop
mass flow rates reach a stable values corresponding to
single phase stable natural circulation. The primary
coolant draining design outlined in Fig. 5 is used in
the calculations documented in this chapter and de-rives from an experiment performed in the LOBI fa-
cility available at the European Community research
center of Ispra in Italy.
Table 5 reports the initial and boundary conditions
of the VVER-1000 natural circulation scenario for our
analysis.
By stepwise decrease of the mass inventory in the
primary side single phase, two phase, reflux con-
densation, and dry-out phenomena or events occur
(Mousavian et al., 2003a, 2003b). The chronological
major events are described in Section 3.1. In the sub-sequent sections the effects of buoyancy force, power
and loop seal on natural circulation are discussed.
3.1. Effect of buoyancy on NC in a VVER-1000
The schematic of a simplified VVER-1000 loop in-
cluding vessel, hot leg, steam generator, cold leg loop
seal and cold leg horizontal part is shown in Fig. 6.
By integrating of one-dimensional momentum equa-
tion in natural circulation conditions (Ppumps = 0)
around in the loop of Fig. 6 we find (Todreas and
Kazimi, 1990):
k
Lk(Gm)k
t= PB + Paccl Pf (7)
Fig. 6. Schematic diagram of a VVER-1000 loop.
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34 S.K. Mousavian et al. / Nuclear Engineering and Design 229 (2004) 2546
where
PB [(m)CL (m)HL]g(Z)CL-HL,
Paccl
k
Gm2
m
k
, and
Pf PRPV + PSGs
+Ppipelines + Ppump-off+ (8)
In Eq. (8) the parameters PB, Paccl and Pf refer
to buoyancy, acceleration and frictional pressure head
terms, respectively. The Eqs. (7) and (8) show the
importance of the term Z. Otherwise, the effect of
acceleration pressure head with respect to other terms
is negligible. Owing to this, the attention is focused
toward the difference in elevation between the heatsink (steam generator in the present case) and the heat
source (core in the present case).
Therefore, natural circulation phenomena are eval-
uated by the help of the RELAP5 code, considering
three different distances (or three elevation differ-
ences) between hot and cold zones in next three
sub-sections (the words high-level and low-level
are used as synonymous and abbreviation for high-
elevation-difference and small-elevation-difference
between heat source and heat sink, respectively).
3.1.1. NC in nominal case
In this sub-section, the natural circulation phenom-
ena are studied in nominal conditions (ZCL-HL =
9.28 m). Variations of fluid and saturation tempera-
tures through the whole loop in natural circulation
conditions (at t = 2000 and t = 10,000 s) can be
found in Fig. 7. The average sub-cooling of the loop
at t = 2000 s can be deduced from Fig. 7a. On the
contrary, the Fig. 7b shows that saturation conditions
occur throughout the loop at t= 10,000 s and differ-
ences shown in the figure give an idea of numericalerrors and of errors (typical) in interpreting the code
results by the code user. Single phase natural circula-
tion occurs between 100 and 82% of primary side mass
inventory. By decreasing mass inventory, at 75% of
nominal conditions, two phase mass flow rate reaches
its maximum value. At low primary coolant mass in-
ventory, i.e., about 40%, the flow rate stagnates.
Fig. 8 shows the variations of pressure in primary
and secondary side reported as a function of total in-
ventory of the primary side. The primary side depres-
surizes slowly and finally reaches the secondary side
pressure and saturation conditions are basically estab-
lished in both primary and secondary sides.
3.1.2. NC in high-level case
By increasing the elevation of steam generators
(heat sinks) with respect to the reactor core (heat
source) the effect of buoyancy force is studied
(ZCL-HL = 15.7 m). The natural circulation mass
flow rate in the loop for both single and two phase
flow regimes is higher (for about 20%) than in the
nominal case. In addition, smaller amplitudes for the
oscillations are calculated, Fig. 9. The conclusion,
as expected is that the cooling capability by natural
circulation is enhanced and is more stable when in-
creasing the elevation difference between heat sink
and heat source.
At very low mass inventories, i.e., coolant mass25% of the nominal value, the dry-out phenomenon
occurs. Fig. 10 shows the consequences of dry-out
upon the clad surface temperature.
3.1.3. NC in low-level case
The height difference between hot and cold zones is
decreased in this case by about two meters with respect
to the nominal case (ZCL-HL = 7.48 m). Core inlet
mass flow rate in single and two phase flow conditionsis calculated to decrease for about 7%. In this case,
two phase flow instability is not observed between
100 and 43% of mass inventory (Fig. 11), with main
reference to the time window around 7000 s.
Considering all the three code runs (nominal,
high-level and low-level), it may be noted that when
low mass inventories (starting from 40% of nominal
value) occur the core inlet mass flow rate is quite sim-
ilar, thus independent upon the elevation difference
between heat source and heat sink.
Table 6 summarizes the chronological events duringnatural circulation in the three analyzed cases. It shall
be noted, as expected, that the event under consider-
ation for the nominal case occurs within the ranges
and/or the occurring times calculated for the high-level
and the low-level cases.
3.2. Effect of loop seal on NC in a VVER-1000
Loop seals are the U-shaped bends in the cold
legs piping connecting steam generators outlet to
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S.K. Mousavian et al. / Nuclear Engineering and Design 229 (2004) 2546 35
At time=2000 s
250
260
270
280
290
300
310
320330
340
350
1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 33
mesh points through first loopFluidandsaturatedtempera
ture(c)
Lower plenum Core region Upper plenum HL SG CL & down
comer
Fluid temperature
Saturated temperature
279.4
279.5
279.6
279.7
279.8
279.9
280
280.1
1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 33
Mesh points through first loop
Fluidandsaturationtemperature(C)
At time=10000 s
(a)
(b)
Fig. 7. Fluid and saturation temperatures through the loop at NC conditions.
MCPs suction. The influence of eliminating the loop
seals upon the natural circulation scenario is dis-
cussed hereafter. This has been done starting from the
VVER-1000 nodalisation representing the real NPP.
The capability of heat removal by natural circulation
in Western PWRs, by using of L-shaped config-
uration instead of U-shaped loop seal, has been
already estimated (DAuria and Frogheri, 2002): such
a capability was calculate to increase for about 20%.
Fig. 12 shows the distribution of liquid velocity
and void fraction through vertical section of cold
leg between steam generator and pump suction that
result from the present VVER-1000 related cal-
culation.
In the case when loop seals are eliminated from the
system configuration, the velocity of liquid after re-
flux condensation is higher than in the original case
(with loop seals). Also during the reflux condensation
mode of natural circulation, the vertical section be-
tween steam generator and pump suction is cleared at
an earlier time in the case when the loop seal is present
compared with the case without loop seal. Moreover,
by eliminating of the loop seals, the instability re-
gion of two phase flow is decreased. Limited or no
effect of the loop seal is detected in other modes of
the natural circulation, thus confirming that removal
of loop seal is beneficial from the point of view of the
thermalhydraulic design of a VVER-1000.
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36 S.K. Mousavian et al. / Nuclear Engineering and Design 229 (2004) 2546
4.00
5.40
6.80
8.20
9.60
11.00
12.40
13.80
15.20
30.040.050.060.070.080.090.0100.0
Inventory of primary side (%)
Primaryandsecondarypress
ure(MPa)
primary
secondary
Fig. 8. Primary and secondary side pressures in nominal case ofNC conditions.
3.3. Effect of core power on NC in a VVER-1000
The core power level is one of the important pa-
rameters affecting the natural circulation scenario in
VVER-1000. Natural circulation experiments from
ISB-VVER (with vertical steam generators) integral
test facility showed that by increasing power from
100 to 200 kW the mass flow rate through the loops
increased by 25% of single phase nominal natural
circulation value (Krepper, 1999).
Therefore, by keeping the reactor power at 2.5, 5
and 10% of nominal power, the cold leg mass flow
-100
0
100
200
300
400
500
600
700
-1000 1000 3000 5000 7000 9000 11000 13000 15000
Time (s)
Massflowt
hrou
ghtheloop(kg/s) Nominal
High-level
Fig. 9. Mass flow rates through the loop in nominal and high level
cases.
20 0
40 0
60 0
80 0
1000
1200
1400
1600
1800
0 5000 10000 15000 20000 25000 30000
Time (s)
Maximumc
laddingtempera
ture(C)
Starting of dry-out
Fig. 10. Maximum cladding temperature in high level case of NC
study (dry-out effect).
rates are obtained (Fig. 13) by using the VVER-1000
nodalisation.
When the power level is decreased from 5 to 2.5%
the core inlet mass flow rate was found to decrease
for about 28% of its nominal value and the two phase
flow instability occurs at a later time (Fig. 13). At
2.5% of nominal power, about 50 minutes after the
natural circulation transient initiation, two phase flow
starts and the void fraction increases up to about 0.3
without dry-out occurrence (till the time when 30% oftotal mass inventory).
When the power level is kept at 5% of the nominal
value, the void fraction is increased and after 180 min
-100
0
100
200
300
400
500
600
700
0 2500 5000 7500 10000 12500 15000
Time (s)
Massflow
rateth
roughtheloop(kg/s)
Fig. 11. Mass flow rate through the loop at low level case of NC
study.
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Table 6
Chronological of events for different cases of natural circulation phenomena
Time (s) Residual mass (%) Events
Nominal conditions (1) High-level conditions (2)
0 t 2800 (1) 82 RM 100 Single-phase flow
(2) 86 RM 100 Single-phase flow
(3) 85 RM 100
t= 3100 (1) RM= 75 Two-phase stable (maximum flow rate)
(2) RM= 77 Two-phase stable (maximum flow rate)
(3) RM= 74
(1) 4000 t 6000 53 RM 58 Two-phase & unstable
(2) 4000 t 4650 57 RM 59 Two-phase and very low unstable
(2) 4000 t 7500 48 RM 57
(1) t= 7850 Reversed flow starts (RPV outlet)
(2) t= 9000 RM= 48
(3) t= 7600 Reversed flow starts (RPV outlet)
(1) 10200 t 11700 40 RM 43 Two-phase with oscillations
(2) 12800 t 14550 38 RM 40 Two-phase with oscillations
(3) 9750 t 10800 41 RM 43
(1) 1170 t 15000 35 RM 40 Reflux condensation (without dry-out)
(2) 14550 t 23000 27 RM 38 Reflux condensation (without dry-out)
(3) 10800 t 15000 34 RM 41
(2) t= 23000 RM= 27 Dry-out starts
(2) 23000 t 25000 24 RM 27 Dry-out and core inlet mass flow rate
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38 S.K. Mousavian et al. / Nuclear Engineering and Design 229 (2004) 2546
-1
0
1
2
3
4
5
6
-500 2500 5500 8500 11500 14500
Time (s)
Liquidvelocity(m/s
)
-0.2
0
0.2
0.4
0.6
0.8
1
1.2
Voidfraction
with loop sealwithout loop sealwithout loop sealwith loop seal
Fig. 12. Loop seal effect on liquid velocity and void fraction.
(or at a value of 38% for total mass inventory) dry-out
occurs.
When the power level is increased from 5 to 10%
of nominal power, about 50 min after the natural cir-
culation transient initiation, two phase flow occurs but
only for 12 min (or 38% of total mass inventory) and
after that the dry-out would be initiated (Fig. 14).
Definitely the increase of power causes:
Larger value for single and stable two-phase natural
circulation, Early appearance of instabilities,
-400
100
600
1100
1600
2100
2600
3100
0 2000 4000 6000 8000 10000 12000 14000 16000
Time (s)
Coreinletm
assflow
rate(kg/s)
(1) power: 300 MW
(2) power: 150 MW(3) power: 75 MW
Single-phase
Two-phase & dry-out (1)
Two-phase (2)Dry-out (2)Two-phase (3)
Two-phase unstable (3)
Two-phase stable (3)
Dry-out (3)
(1), (2) and (3)
Fig. 13. Core inlet mass flow rates at three different power levels.
0
0.2
0.4
0.6
0.8
1
-500 1000 2500 4000 5500 7000 8500 10000 11500
Time (s)
Voidfraction
power 2.5%
power 5%
power 10%
Fig. 14. Void fraction through reactor core at three different powers.
Anticipation of main events with main reference to
the time when dry-out is calculated.
4. Natural circulation performance using
NCFM
One of the methods for evaluating the natural circu-
lation phenomena is the natural circulation flow map
(NCFM), as already mentioned. Core inlet mass flow
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S.K. Mousavian et al. / Nuclear Engineering and Design 229 (2004) 2546 39
-5
0
5
10
15
20
25
30
35
40
200 300 400 500 600 700
Residual mass per volume (kg/m^3)
Massflowrateperpower(Kg/s-M
Wth)
NominalLow-levelHigh-level
Fig. 15. NCFM at different heights (buoyancy effect) of a
VVER-1000.
rate, reactor core power, total mass inventory and vol-
ume of primary side are considered within the NCFM
approach to evaluate the natural circulation perfor-
mance of systems (e.g., including code nodalisation
representative of natural circulation systems) not used
to construct the NCFM itself. A NCFM has been ob-
tained by the envelope of experimental data (lower
and upper limits) from several commercial NPPs andintegral test facilities such as Bethsy, Lobi, Semis-
cale, Lstf, PKL and SPES (DAuria and Frogheri,
2002).
The natural circulation performance of VVER-1000
at three different heights can be derived from Fig. 15
where the NCFM is not reported. It clearly ap-
pears that the natural circulation performance of
the high-level design is better than the two other
designs.
The NCFM is used in Fig. 16 for comparing the
natural circulation performance of VVER-1000 withthe performance of Western PWRs integral test facili-
ties.
The maximum values of mass flow rate (per unit
volume of primary system) for VVER-1000 and PWR
NPPs are similar. However, in VVER-1000, related to
Western PWRs, the mass flow rates through loop re-
main stable and above the zero value (characteristic of
the reflux condensation in the case of PWR) at low val-
ues of primary system inventory. Thus, VVER-1000
shows a better NCFM than PWR.
-5
0
5
10
15
20
25
30
35
40
200 300 400 500 600 700 800 900 1000
Residual mass per volume (kg/m^3)
Massflowrateperpower(kg/s-MWth)
lower-limit (PWR)
upper-limit (PWR)
nominal (VVER-1000)
high-level (VVER-1000)
Fig. 16. Comparison of NCFM for VVER-1000 in nominal and
high level conditions with lower and upper limits in Western PWR.
5. Natural circulation relevant to SBLOCA
A series of SBLOCA analyses with break area rang-
ing from 0.5% (5 mm diameter) to 11.7% (100 mm
diameter) of the VVER-1000 cold leg diameter are
performed starting from NPP nominal conditions. The
initial and boundary conditions are summarized in
Table 7. The break is placed in the cold leg of theloop with pressurizer between the pump outlet and the
down-comer.
Generally during SBLOCA in Western PWRs there
are five different phases (Prosek and Mavko, 1999) in-
cluding sub-cooled blow-down, two-phase natural cir-
culation, reflux condensation, loop seal clearing, core
re-flooding, and long term cool-down.
In the SBLOCA experiments, in PACTEL facility
(medium scale facility of VVER-440), flow stagna-
tion and system re-pressurization were observed when
the water level in the upper plenum fell below the en-trances to the hot legs at a time when coolant flow into
the hot legs changed from single to two-phase flow
(Lomperski and Kouhia, 1994). The core inlet mass
flow rate for different break sizes is depicted in Fig. 17.
In the SBLOCA with 11.7 and 5.8% break sizes,
primary side pressure decreases and reaches secondary
side pressure at 75 and 425 s, respectively (Fig. 18).
However, in different SBLOCA scenarios, primary
side pressure was predicted to remain higher than sec-
ondary pressure.
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40 S.K. Mousavian et al. / Nuclear Engineering and Design 229 (2004) 2546
Table 7
Initial and boundary conditions in NC relevant to SBLOCA
Initial conditions Boundary conditions
Parameters Unit Values Parameters/systems Values/conditions
Reactor power MW 3100 Time of trip and reactor scram t= 0 s
Primary/secondary sides pressures MPa 15.7/6.2
Mass inventory of RPV/primary-side kg 81.7 103/232.6 103 ECCS (LPIS, HPIS, ACCU) Activated
Fluid temperature at RPV inlet/outlet C 287.0/322.7 Auxiliary/emergency FW Not activated
Fluid temperature at SG inlet/outlet C 322.7/287.0 Main FWs Activated
PRZ temperature C 345.8 MSIV valves Closed (after transient)
Fuel/cladding maximum temperatures C 2200/351.5 PORV valves Not activated
Collapsed level of PRZ m 8.5
Collapsed level of core m 3.25 Break valve opening time t= 0 s
Collapsed level of SG m 1.9
FWs temperatures C 220.3 Break area (%) 0.5, 1.1, 3, 5.8 and 11.7
0 .2 .4 .6 .8 1 1.2 1.4 1.6
X 104Time (s)
-200
0
200
400
600
800
1000
1200
1400
1600
Core
inletmassflow
rate(kg/s)
XXX 11.7%-with-ECCSX
X
X
X X X
YYY 5.8%-with-ECCS
Y
Y
Y
YY Y
ZZZ 5.8%-without-ECCS
Z
Z
Z
Z
Z Z
0
200
400
600
800
1000
1200
1400
1600
1800
0.0E+00 2.0E+03 4.0E+03 6.0E+03 8.0E+03 1.0E+04 1.2E+04 1.4E+04 1.6E+04
Time (s)
Coreinletmassflowrate(kg/s)
0.5%-with-ECCS
1.1%-with-ECCS
1.1%-without-ECCS
3.0%-with-ECCS
3.0%-without-ECCS
(a)
(b)
Fig. 17. Core inlet mass flow rate in SBLOCA at different breaks
and conditions.
SBLOCA experiments by using ECCS on the
ISB-VVER integral test facility showed that the core
collapsed level and the primary pressure decrease
slower than in the case without using emergency
core cooling system. Also, the maximum cladding
temperatures do not undergo to dangerous values
(Krepper, 1997). In the SBLOCA (5.8% break) with-
out ECCS (including HPIS, LPIS and ACCU), at
less than 26% of total mass inventory, dry-out oc-
curs. Fig. 19 shows the gas void fraction through
reactor core when different breaks are assumed tooccur.
In the study of SBLOCA (break area less than 3%
of the pipeline area) with actuation of ECCS, void
fraction through core is zero. But in the case of without
ECCS at 3% and also1.1% void appear (Fig. 19), as
expected.
The comparison between SBLOCA (5.8% break
area with and without ECCS) and Natural circula-
tion scenarios (with draining method) is shown in
Fig. 20 in the time domain even though the repre-
sentation in the phase-space (i.e., with coolant inven-tory on the horizontal axis and pressure in the vertical
axis) could be more representative. The main results
of comparison showed that, during the SBLOCA with
ECCS:
Instability of two-phase flow is not observed; The maximum value of mass flow rate is decreased
for about 30%.
Both the two differences between natural circulation
and SBLOCA scenarios are connected with the exis-
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S.K. Mousavian et al. / Nuclear Engineering and Design 229 (2004) 2546 41
Fig. 18. Pressure distribution of primary and secondary side in different SBLOCAs.
tence of the driving-force represented by the pres-
sure difference across the break that is not present in
the case of the natural circulation.The SBLOCA study confirmed that some safety
margins, e.g., time of dry-out occurrence, Figs. 18c
and 19, increases more than eight times (see Fig. 19)
in the cases when ECCS are used, related to cases
when ECCS are not part of the system design.
Definitely, the performed SBLOCA analyses
showed that natural circulation is hugely affected
by the selected sequence of events where break area
and actuation modes of emergency system play the
greatest role.
5.1. Reflux condensation cooling in VVER-1000
As the mass inventory of the primary side is de-creased, the natural circulation is terminated and the
steam begins to be condensed in steam generators
tubes (starting point is the hot packages of upper pipes)
and propagates through hot legs, see Fig. 21, and the
reactor pressure vessel (Mousavian et al., 2003a). The
propagation of the discontinuity in the liquid velocity
(Fig. 21) is actually representative of the propagation
of void fraction into the cold leg.
Table 8 reports some information related to the start-
ing of flow reversal through hot leg at different posi-
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42 S.K. Mousavian et al. / Nuclear Engineering and Design 229 (2004) 2546
Fig. 19. Gas void fraction through core in different SBLOCAs.
tions along the hot leg itself (these are characterized
by a node number pertaining to the nodalisation) based
on mass inventory. When the collapsed liquid level in
the reactor pressure vessel goes below the level of the
Table 8
Starting point of reflux condensation between RPV outlet and ascending side of SG
Residual mass (%) Component (relevant to Fig. 1) Nominal case High-level case Low-level case
38.0 201-01 Reversed flow
41.1 200-01 Reversed flow
41.7 200-01 Reversed flow
46.3 210-01 Reversed flow
46.7 210-01 Reversed flow
47.0 210-01 Reversed flow
Fig. 20. Comparison between SBLOCA (5.8%) and NC-with-
draining in a VVER-1000.
hot leg (entrance) there is the potential for flow rever-
sal (Figs. 21 and 22). Looking at Fig. 22, this occurs
at a time bounded by 3000 s and 12,000 s in the con-
sidered SBLOCA scenario.
When the reactor cools down by reflux condensation
the counter current flow limitation (CCFL) or flood-
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S.K. Mousavian et al. / Nuclear Engineering and Design 229 (2004) 2546 43
-3
-2
-1
0
1
2
3
4
5
6
-500 1500 3500 5500 7500 9500 11500 13500 15500
Time (s)
Liquidvelocitythroughhotlegfrom
RPV
toSG
(m/s)
comp. 200
comp. 205
comp. 210
Fig. 21. Liquid velocity through hot leg from RPV to SG.
ing could occur. The supply of cooling liquid (i.e.,
from the ascending side of steam generators, namely
of PWR type NPP) into the reactor core is limited by
the occurrence of CCFL at the steam generators inlet
plenum (namely, in PWR types NPP) or in the connec-
tion between hot legs and reactor pressure vessel (this
could be of some relevance in both PWR and VVER).
The potential effect of reactor power upon reflux
condensation can be envisaged from observing thetime trends in Fig. 23. By increasing the reactor
power from 75 to 150 MWth the time when negative
liquid flow-rate is calculated for the horizontal tubes
of steam generators occurs earlier with value of mass
7.0
8.0
9.0
10.0
11.0
12.0
13.0
0.0E+00 3.0E+03 6.0E+03 9.0E+03 1.2E+04 1.5E+04
Time (sec.)
Collapsedlevelofliquid
inreactorpressure
vessel
(m)
Fig. 22. Reactor pressure vessel collapsed level at nominal
conditions.
0 .2 .4 .6 .8 1 1.2 1.4 1.6
X 104
Time (s)
-.40
-.20
0
.20
.40
.60
.80
LiquidvelocitythroughhotpackageofS
G1(m/s)
XXX power:75MW
X X X
X
XX
X
XX
XX
X
X X X X X X X
YYY power:150MW
Y Y Y YY
Y
Y
YY Y
Y
Y
Y
Y
Y Y Y Y
ZZZ power:300MW
Z
ZZZZZZZZZZZZ
Z
Z
Z
ZZ
Z
Fig. 23. Liquid velocity through the pipes of steam generator at
different power levels.
inventory that changes from 30 to 40% of the initial
value. It should be noted that negative value for liquid
velocity does not imply directly the occurrence of
CCFL but gives an idea for the potential occurrence of
such phenomenon. The search for CCFL conditions
should be based upon specific transient calculation
of licensing or design interest that was not within thepurpose of the present activity.
As mentioned above, the reflux condensation phe-
nomenon is one of the effective heat removal mech-
anisms during SBLOCA (Kawanishi et al., 1991)
in PWR. The break size effect on reflux condensa-
tion is important, as confirmed from the present ac-
tivity.
From the analyses of VVER-1000 SBLOCA tran-
sients originated by break areas between 11.6 and
5.8%, the reflux condensation phenomenon is not ob-
served. The ultimate reason is that in those conditionsthe heat exchange between primary and secondary
sides of SGs is lower than the thermal power removal
via the break. However, in the calculation of SBLOCA
originated by a 3% area, condensation in the upper
rows of horizontal tubes is calculated.
From the calculation of the SBLOCA scenario
originated by the 1.1% break area, flow reversal and,
therefore the potential for reflux condensation, ap-
pears evident at the time when 60% of mass inventory
is observed. Fig. 24 shows the liquid and gas velocity
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44 S.K. Mousavian et al. / Nuclear Engineering and Design 229 (2004) 2546
0 .5 1 1.5 2 2.5 3 3.5
X 104Time (s)
-.20
0
.20
.40
.60
.80
1.00
1.20
1.40
Liquidandgasvelocity(m/s)
XXX Liquid
X
XX X
X
XX X X X X X X
YYY Gas
Y
YY Y
Y
Y
Y
Y
Y
Y
Y
YY
effect of reflux condensation
Fig. 24. Liquid and gas velocity at pipes of steam generator in a
VVER-1000.
in the inlet side of the upper row of horizontal tubes.
Negative values for liquid velocity are indication for
the potential occurrence of the reflux condensation
phenomenon.
6. Conclusions
A wide range investigation of natural circulation
in VVER-1000 has been performed within a devoted
research activity documented in the present paper. This
included:
The use of experimental data (e.g., derived from
PSB and ISB facilities);
The use of the RELAP5 system code;
The use of a qualified VVER-1000 nodalisation and
of procedures to attain the qualification;
The use of the NPP plant data (in the process ofnodalisation qualification);
The use of the (fast Fourier transform based method)
FFTBM to quantify the accuracy or the error that
characterizes the comparison between predicted and
measured sets of data;
The use of the NCFM to evaluate the natural circu-
lation performance of the VVER-1000 NPP.
Selected thermalhydraulic phenomena expected in
case of natural circulation in VVER-1000 geometry
are studied by considering both fluid draining from
the cold leg and SBLOCA scenarios.
The consideration of buoyancy effect showed that
by increasing (or decreasing) the average distancebetween the heat sink and the heat source, consti-
tuted by the steam generator tubes and the core in
the VVER-1000 loop respectively, the mass flow rate
through the loop is increased (or decreased). It has
also been calculated that the stability performance
of the entire systems is also substantially affected
by this geometric parameter. The core power effect
upon an established natural circulation scenario has
also been characterized and compared with the ef-
fect derived in the case of Western PWRs, with main
reference to the single-phase regions (DAuria et al.,
1991).The influence of the presence of the loop seal, i.e.,
the U-shaped pipe between the steam generator outlet
and the pump suction in the primary loop has been in-
vestigated. Its noticeable influence upon both the natu-
ral circulation flow-rate and the stability performance
of the system has been quantified.
The application of the NCFM showed that the nat-
ural circulation performance of the VVER-1000 nu-
clear power plants is equivalent or better than the nat-
ural circulation performance of Western PWRs, espe-
cially for small values of the primary system massinventory.
Natural circulation in VVER-1000 systems has also
been analysed during SBLOCA transients originated
by different break area. The effects of ECCS including
HPIS, LPIS and accumulators on the SBLOCA have
been evaluated and their relevance has been character-
ized in quantitative terms. In case of SBLOCA, it was
found that the overall system is less prone to insta-
bility than in the case when the coolant draining pro-
cedure is applied. This derives from the presence of
the driving force in case of SBLOCA scenarios origi-nated by the pressure difference across the break. The
presence of emergency systems also makes the sys-
tem more stable with respect to the case when those
systems are not present.
The potential for the occurrence of reflux conden-
sation in VVER-1000 conditions has been studied. At
a preliminary level it was found that conditions for
reflux condensations are not established when break
area has a value larger than 3% of the area of the cold
leg.
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S.K. Mousavian et al. / Nuclear Engineering and Design 229 (2004) 2546 45
Acknowledgements
The authors gratefully thank to Dr. G.M. Galassi
(University of Pisa, Italy) and Dr. N. Fil (Gidropress,Russia) for the review of the paper and the fruitful
discussions had when performing the present activity.
One of the authors (S.K. Mousavian) appreciates the
Ministry of Sciences, Researches and Technology of
Iran and to the AEOI for supporting of this work.
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