# Use of the Natural Circulation Flow Map for Natural Circulation Systems Evaluation

Authors: M. Cherubini; W. Giannotti; D. Araneo; F. D'Auria
Journal: Science and Technology of Nuclear Installations
 (2008)

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## Abstract

The aim of this paper is to collect and resume the work done to build and develop,
at the University of Pisa, an engineering tool related to the natural circulation. After
a brief description of the different loop flow regimes in single phase and two phase,
the derivation of a suitable tool to judge the NC performance in a generic system is
presented. Finally, an extensive comparison among the NC performance of various
nuclear power plants having different design is done to show a practical application of the
NC flow map.


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## Body

## 1. Introduction

The natural circulation (NC) is an important mechanism in several
industrial systems and the knowledge of its behaviour is of interest to nuclear
reactor design,
operation, and safety. In the nuclear technology,
this is especially true for new concepts that largely exploit the gravity
forces for the heat removal capability. The NC takes place due to the presence
of a heat source and of a heat sink constituted in a nuclear power plant (NPP) by
the core and the steam generators, respectively. In a gravity environment, with
the core located at a lower elevation than the steam generators, driving forces
occur such that to generate suitable flow rate for removing nuclear fission
power. While the NC core power removal capability was only exploited for
accident situations basically to demonstrate the inherent safety features of
the plants, recently within the passive reactor concept the NC performance assumes
major importance having a large impact on the design phase.The NC behaviour
has been (and it is still now) object of several experiments aimed at addressing
issues like scaling or characterization test in a test rig. In this view, a
quite large database has been created collecting the experimental measurements from
various facilities. The NC scenarios occurring for different values of the
primary system mass inventory were considered (reference is made to both single-phase
and two-phase natural circulation) by gathering and analyzing data from the following
PWR simulators: Semiscale, Spes, Lobi, Bethsy, Pkl, and Lstf [1].In order to
evaluate the natural circulation performance (NCP) of the mentioned facilities,
significant information comes from the analysis of the trend of the core inlet
mass flow rate and the primary loop mass inventory. The flow rate and the
residual masses have been normalised taking into account the volume of each
facility and the corresponding power level (typically ranging between 1 and 5%
of the nominal core power) utilized in the selected experiment. Four main flow
patterns were characterized depending upon the value of the mass inventory of
the primary loop.


## 2. The Natural Circulation Flow Regimes

### 2.1. Single-Phase NC (SPNC)

SPNC regime implies no void occurrence in the upper plenum of the
system. Therefore, coolant at the core outlet will be subcooled up to nearly
saturated. Core flow rate is derived from the balance between driving and
resistant forces. Driving forces are the result of fluid density differences
occurring between{descending side of U-tubes and vessel downcomer} and {core and ascending side of U-tubes}. Resistant
forces are due to irreversible friction pressure drops along the entire loop.
Resulting fluid velocities are sufficient for removing core power in
(subcooled) nucleate boiling or forced convection heat transfer regimes: no
film boiling condition is experienced in the core. It may be noted that the
secondary side of SG is also a natural circulation system working in two-phase
conditions.SPNC may occur at any primary system pressure, consistently with SG
pressure. However, typical primary system pressures range between 8 and 16 MPa
with secondary pressure close to the nominal operating condition. It is
expected from the NPP design that SPNC, provided the availability of SG
cooling, is capable to remove the nuclear decay heat from the core.
Experimental database, including NPP tests, confirms this capability.


### 2.2. Stable Two-Phase NC (TPNC)

TPNC regime occurs as a consequence of coolant loss from the primary
system. Owing to this, both driving and resistant forces increase when
decreasing mass inventory of primary system. Assigned the typical geometrical
layout of PWR, the former effect, that is, increase of driving forces, is
prevalent at small decreases of mass inventories. The opposite occurs for larger
decreases of mass inventories. The net result is a “peak” in core mass flow rate
versus primary system inventory (when primary mass flow rate decreases) as can
be observed in Figure1. Forced convection, subcooled, and saturated heat
transfer regimes occur in the core. Condensation occurs inside the U-tubes of
SG. The average core void fraction is typically less than 30%, whereas at the
outlet values around 50% can be reached without occurrence of thermal crisis in
the considered pressure range.Figure 1
Characterization of natural circulation flow
regimes based on experimental data and system code calculations.


### 2.3. Siphon Condensation NC (SCNC)

The decreasing of NC driving forces, the small temperature
difference across U-tubes of the steam generators, and the occurrence of the countercurrent
flow limiting phenomenon (CCFL) at the entrance of U-tubes (e.g., see [2]) are
at the origin of wide system oscillations of core inlet flow rate. This
phenomenon has been investigated in [3, 4] and based on a natural circulation
experiment performed in Lobi facility in [5]. However, evidence of the siphon
condensation has been found also in other facilities.At mass inventories of the primary system around 70% of the nominal
value, the efficiency of the condensation heat transfer across U-tubes causes
the release of almost all core thermal power in the ascending side of U-tubes. Liquid
level builds up and is prevented to drain down by the steam-liquid mixture
velocity at the tube entrance, that is, the CCFL condition occurs. Therefore,
liquid level rises in the U-tubes till reaching the top. During this period,
typical duration of the order of 10 seconds, the flow rate at the core inlet is
close to zero hence the core boiloff occurs. Once the liquid level reaches the
upper bend of U-tubes, the siphon effect occurs and causes the emptying of the
ascending side of U-tubes and the reestablishment of core inlet flow rate. A
new cycle starts. The phenomenon is made more complex by the interaction of the
several thousands of U-tubes that constitute a SG tube bundle. Different groups
of tubes may stay at a different stage of the oscillation at the same time,
also causing flow reversal in the tube bundle. Suitable core cooling still can
be achieved in these conditions.


### 2.4. Reflux Condensation (RCNC)

At “low” mass inventories of primary coolant and/or at low core
power, the steam velocity in the upper part of the system including hot legs
and steam generator entrance is low. Weak interactions occur at the
steam-liquid interface that are not enough to cause CCFL. In these conditions,
the liquid that is condensed or entrained in the ascending side of the U-tubes may
flow back to the hot leg and to the core. Stratified countercurrent steam and
liquid flow simultaneously in the hot legs. The mass flow rate at the core
inlet is close to zero, although a “minor” natural circulation path may
establish between core and downcomer inside the vessel. However, upward
two-phase mixture and downward liquid flows occur at the core outlet. Core
thermal power can be removed by boiloff in the saturated nucleate boiling heat
transfer regime.


### 2.5. Dryout Occurrence

The terms “dryout occurrence” appear in the right part of Figure1, when primary
system mass inventory is roughly lower than 40% of the nominal value. Dryout is
caused by the combination of low flow and high void fraction. As a consequence,
film boiling heat transfer regime is experienced with low coefficient for heat
transfer. Rod surface temperature increases in various zones of the core, and
the overall process of thermal power transfer from fuel rods to the fluid may
become unstable. The system operation in these conditions is not acceptable
from a technological point of view. It can be noted that the temperature
excursion is strongly affected by primary system pressure and thermal power
levels: the linear rod power plays a role in these conditions. At primary
system pressure around 15 MPa (nominal operation for PWR), “post-dryout”
surface temperature jumps may be as low as a few tens of Kelvin,
tolerable for the mechanical resistance of the rod-clad material.All the regimes cited above are summarized in Figure1 which reports
both experimental and calculated data.
On the vertical axis the percentage of the nominal core
power (P) is reported, while the residual mass inventory in the primary circuit
(RM) is reported on the horizontal axis.


## 2.1. Single-Phase NC (SPNC)

SPNC regime implies no void occurrence in the upper plenum of the
system. Therefore, coolant at the core outlet will be subcooled up to nearly
saturated. Core flow rate is derived from the balance between driving and
resistant forces. Driving forces are the result of fluid density differences
occurring between{descending side of U-tubes and vessel downcomer} and {core and ascending side of U-tubes}. Resistant
forces are due to irreversible friction pressure drops along the entire loop.
Resulting fluid velocities are sufficient for removing core power in
(subcooled) nucleate boiling or forced convection heat transfer regimes: no
film boiling condition is experienced in the core. It may be noted that the
secondary side of SG is also a natural circulation system working in two-phase
conditions.SPNC may occur at any primary system pressure, consistently with SG
pressure. However, typical primary system pressures range between 8 and 16 MPa
with secondary pressure close to the nominal operating condition. It is
expected from the NPP design that SPNC, provided the availability of SG
cooling, is capable to remove the nuclear decay heat from the core.
Experimental database, including NPP tests, confirms this capability.


## 2.2. Stable Two-Phase NC (TPNC)

TPNC regime occurs as a consequence of coolant loss from the primary
system. Owing to this, both driving and resistant forces increase when
decreasing mass inventory of primary system. Assigned the typical geometrical
layout of PWR, the former effect, that is, increase of driving forces, is
prevalent at small decreases of mass inventories. The opposite occurs for larger
decreases of mass inventories. The net result is a “peak” in core mass flow rate
versus primary system inventory (when primary mass flow rate decreases) as can
be observed in Figure1. Forced convection, subcooled, and saturated heat
transfer regimes occur in the core. Condensation occurs inside the U-tubes of
SG. The average core void fraction is typically less than 30%, whereas at the
outlet values around 50% can be reached without occurrence of thermal crisis in
the considered pressure range.Figure 1
Characterization of natural circulation flow
regimes based on experimental data and system code calculations.


## 2.3. Siphon Condensation NC (SCNC)

The decreasing of NC driving forces, the small temperature
difference across U-tubes of the steam generators, and the occurrence of the countercurrent
flow limiting phenomenon (CCFL) at the entrance of U-tubes (e.g., see [2]) are
at the origin of wide system oscillations of core inlet flow rate. This
phenomenon has been investigated in [3, 4] and based on a natural circulation
experiment performed in Lobi facility in [5]. However, evidence of the siphon
condensation has been found also in other facilities.At mass inventories of the primary system around 70% of the nominal
value, the efficiency of the condensation heat transfer across U-tubes causes
the release of almost all core thermal power in the ascending side of U-tubes. Liquid
level builds up and is prevented to drain down by the steam-liquid mixture
velocity at the tube entrance, that is, the CCFL condition occurs. Therefore,
liquid level rises in the U-tubes till reaching the top. During this period,
typical duration of the order of 10 seconds, the flow rate at the core inlet is
close to zero hence the core boiloff occurs. Once the liquid level reaches the
upper bend of U-tubes, the siphon effect occurs and causes the emptying of the
ascending side of U-tubes and the reestablishment of core inlet flow rate. A
new cycle starts. The phenomenon is made more complex by the interaction of the
several thousands of U-tubes that constitute a SG tube bundle. Different groups
of tubes may stay at a different stage of the oscillation at the same time,
also causing flow reversal in the tube bundle. Suitable core cooling still can
be achieved in these conditions.


## 2.4. Reflux Condensation (RCNC)

At “low” mass inventories of primary coolant and/or at low core
power, the steam velocity in the upper part of the system including hot legs
and steam generator entrance is low. Weak interactions occur at the
steam-liquid interface that are not enough to cause CCFL. In these conditions,
the liquid that is condensed or entrained in the ascending side of the U-tubes may
flow back to the hot leg and to the core. Stratified countercurrent steam and
liquid flow simultaneously in the hot legs. The mass flow rate at the core
inlet is close to zero, although a “minor” natural circulation path may
establish between core and downcomer inside the vessel. However, upward
two-phase mixture and downward liquid flows occur at the core outlet. Core
thermal power can be removed by boiloff in the saturated nucleate boiling heat
transfer regime.


## 2.5. Dryout Occurrence

The terms “dryout occurrence” appear in the right part of Figure1, when primary
system mass inventory is roughly lower than 40% of the nominal value. Dryout is
caused by the combination of low flow and high void fraction. As a consequence,
film boiling heat transfer regime is experienced with low coefficient for heat
transfer. Rod surface temperature increases in various zones of the core, and
the overall process of thermal power transfer from fuel rods to the fluid may
become unstable. The system operation in these conditions is not acceptable
from a technological point of view. It can be noted that the temperature
excursion is strongly affected by primary system pressure and thermal power
levels: the linear rod power plays a role in these conditions. At primary
system pressure around 15 MPa (nominal operation for PWR), “post-dryout”
surface temperature jumps may be as low as a few tens of Kelvin,
tolerable for the mechanical resistance of the rod-clad material.All the regimes cited above are summarized in Figure1 which reports
both experimental and calculated data.
On the vertical axis the percentage of the nominal core
power (P) is reported, while the residual mass inventory in the primary circuit
(RM) is reported on the horizontal axis.


## 3. The Natural Circulation Flow Map

The database gathered from ten experiments performed in the six integral
test facilities (ITFs) listed in Table1 has been used ([6, 7]) to establish a natural circulation flow map.Table 1
Relevant hardware characteristic of the PWR
simulators considered for the NCFM creation.
Item123456Semiscale Mod2ALobi Mod2SpesPKL-IIIBethsyLstfReference reactor and
power (MWt)W-PWR 3411KWU-PWR 3900W-PWR 2775KWU-PWR 3900FRA-PWR 2775W-PWR 3423No. of fuel rod simulators2564973404281064No. of U-tubes per SG2/68/2413/13/1330/30/6034/34/34141/141Internal diameter of
U-tubes (mm)19.719.615.410.019.719.6Actual Kv1/19571/5891/6111/1591/1321/48In all the considered ITF, NC
experiments with similar
modalities have been performed, and the NC regimes discussed in the previous
chapter are experienced. The linear power of the fuel rod simulators, the
fraction of nominal core power, and the primary system pressure constitute the
main differences for the boundary conditions of the considered experiments. In
relation to the primary side pressure, PKL experiments have been performed at a
pressure value roughly half of the value adopted in the other facilities. The
range of design and operational parameters of the ITF (e.g., pipe diameter,
system volume, number of the steam generators, heat losses to the environment),
not explicitly discussed here, and the identified differences are assumed to
produce an envelope for any expected NC situation in a typical PWR when decay
heat removal is concerned.Measured values of core inlet flow rate (G, kg/s), core power (P,
MW), primary system fluid mass inventory (RM, kg), and net volume of the
primary system (V = const., m3) have been used for setting up the natural
circulation flow map (NCFM). The diagram G/P versus RM/V has been preferred for
the NCFM over other possible choices including nondimensional quantities.The experimental database from ITF (six ITFs, ten experiments) and
the envelope of the curves are given in Figures2 and 3, respectively. The
envelope in Figure 3 is assumed to constitute the NCFM of PWR at core decay power.Figure 2
Natural circulation system behaviour measured
in ten experiments performed in six PWR simulators.Figure 3
Natural circulation flow map achieved from the envelope of measured curves in PWR simulator.A first demonstration of the use of the NCFM has been done in [8],
where seven commercial NPP systems and three ITFs, not used for setting up the
database presented in Figure 2, have been analyzed. Main characteristics of the
NPP and ITF can be drawn from Tables 2 and 3, respectively. The considered NPP
include U-tube, once through and horizontal SG design. The ITFs are Pactel and
RD14M (see Table 3) experimental simulators of WWER-440 and CANDU NPP,
respectively. Their geometric layout is different from those of a PWR. In the
case of WWER-440, six loops equipped with HTSG are connected to the vessel,
though only three are simulated in Pactel.
Horizontal core configuration characterizes the CANDU design, that is, equipped with UTSG.Table 2
Relevant
characteristics of NPP considered for the application of the NCFM.
PWRPWRPWRWWER-1000EPRAP-600EP-1000Nominal power (MWt)187787027333000425019722958Primary system volume (m3)167150330359459211339SG typeU-tubesU-tubesOnce-throughHorizontalU-tubesU-tubesU-tubesNo. of loops2424423No. of pumps2444446Nominal mass inventory (Mg)108108224240307145227Nominal core flow (kg/s)90373150171381528120713826414507Pressurizer and SG
pressure (MPa)15.6 6.14.0 3.115.0 6.415.7 6.315.5 7.215.5 5.515.8 6.4Table 3
PACTEL and RD14M.
Pactel (original)Pactel (with CMT)°RD14MReference reactor and
power (MWt)WWER-440 1375WWER-440 1375CANDU 1800No. of rods14414470No. of SG332SG typeHorizontalHorizontalU-tubesActual Kv+1/4331/4621/378° CMT = core makeup tank.+ Definition introduced for database in Table 1.Comparing the calculated data with the NCFM, authors concluded
that(i)
PWR equipped by
OTSG have poor natural circulation performance;(ii)
NPP designed
around passive system concept showed a quite extend SPNC even at low mass
inventory;(iii)
Russian design reactors equipped by horizontal SG are also suitable for NC mode;(iv)
RD-14m database was
not fully qualified (e.g., the documentation was not so exhaustive), this
brought to a large deviation from the map.The NCFM was used to characterize the behaviour of the CNA-I PHWR
NPP in NC flow conditions, in a reduced primary mass inventory scenario [6].
The simulations have been performed using three different nodalizations of increasing
detail. The coarser one (SET I) was used as a basic set just to represent a
plant layout similar to a PWR (see Figure 4). This permitted to verify that the
trends known for most ITFs working in similar situations give an envelope to
the CNA-I behaviour, considering appropriate trips of some of its safety
systems.Figure 4
Natural circulation map for Atucha-I at different core power level and in case of SBLOCA.


## 4. Recent Use of the NC Flow Map

### 4.1. Application to the PSB-VVER Facility

As already showed by the previous examples (other can be got from, e.g., [8, 9]), the NCFM constitutes an advantageous tool to depict the NC
capability in removing the core power.  Starting from this point more applications have been done that cover
practically all the NPP types.In Figure5 (taken from [10]) it can be seen
the use of the NCFM for a VVER-1000 simulator whose scale factor is 1 : 300. The
experiment has been carried out in the frame of an OECD project in which the
core power and the SG pressure are maintained constant, while the mass is
stepwise drained. Both measured (in blue) and calculated date (in violet) are
reported. The experimental data goes out from the map because the flow rate
measurement systems were not qualified in two-phase flow giving a wrong
indication. However, the calculation confirms on one hand the suitability of
the NCFM and on the other hand the good NC performance of the VVER reactor
type.Figure 5
Natural circulation map for the PSB-VVER experimental and calculated data.


### 4.2. Revisiting the RD-14m Facility

In Section3, the RD14m data were considered but achieving results
far from what expected by the NCFM. Recently, a calculation has been repeated
fixing constant the core power and the SG pressure, while the primary side mass
is stepwise reduced, following a typical way to perform such kind of NC test.
The nodalization had been qualified against LBLOCA test (B-9401) carried out in
the framework of an IAEA project [11]. Analyzing the code results (summarized
in Figures 6 and 7), two main conclusions can be drawn.
(i)
The CANDU facility seems not so suitable for NC status (Figure6). However it should be
noted that the NCFM reasonably resumes the SG NC, it does not consider any
other mechanism simply because they are outside of its definition.(ii)
Figure7 shows
that in the CANDU installation other type of NC are present. The channel to
channel mode seems relevant from the core power removal point of view. This
behaviour explains why even at value of RM/V greater than 500 kg/m3,
where the NCFM predicts low flow condition (hence degraded capability in
removing the core power) the dryout is not experienced.Regarding the latter point (i.e., the channel to channel NC mode), Figure 7 reports the ratio channel mass flow rate over power versus time of two
different horizontal channels. It can be seen that once one core channel has
positive mass flow rate (i.e., the water follows the normal flow path entering
from the inlet and going out from the outlet of the channel), the other is
cooled by water circulating in the opposite path. This means that only a part
of the total core flow rate (sum of the core channels mass flow rate) reaches
the SG to be cooled.Figure 6
Natural circulation map for the RD-14m calculated data.Figure 7
Channel-to-channel natural circulation mode for the RD-14m calculated data.


### 4.3. Consideration of BWR Data

The last use of the NCFM regards the
consideration of boiling water reactor type coming from plant measurement (BWR)
and calculation results (RBMK).Figure8 reports the typical power-flow map of
a BWR in which the NC curve is marked. The maximum removable power in NC mode
is roughly 40% with a core flow rate of 25%.Figure 8
Typical BWR power-flow map.In Figure9 the BWR data are reported into the
NCFM (green dots and green square). Four points are present in the map: full
power condition (green square) at full pump speed, 25%, 4%, and 3% of the
nominal power (green dots). It can be seen that the points are practically
aligned along a line that marks the RM/V of the considered plant confirming the
suitability of the BWR to work in NC mode. Values of G/P comparable to what measured
in TP-flow conditions in PWR simulators are experienced in BWR but with a RM/V
ratio almost halved. The degradation of the performance in core removable power
typical of PWR is not present in BWR that in turns can be operated at 25% of its
nominal power.Figure 9
BWR and RBMK data and comparison with the NCFM.Again in Figure9 it can be seen also the RBMK data.  Those calculated values (red and blue stars are obtained by the use of a
qualified nodalization developed and tested during an EC funded project [12]. The
following consideration come out when calculated data are put inside the NCFM
and compared with BWR values.(i)
The RBMK data stay almost aligned at various power levels as occur in the BWR.(ii)
The full power point of both boiling water reactors (western and eastern
project type, resp.) has the (practically) same G/P ratio, but differs
by a factor of two in the RM/V ratio. This stresses the diverse project type
(vessel type versus channel type reactor), the different nominal point of work,
and the impact on the primary side (PS) mass of the two steam drums used in the
RBMK as separator.(iii)
The G/P ratio remains comparable between the two reactors at various power levels up to 4%.(iv)
Deviation from the consideration of the item above is experienced, when
the bypass between the pressure header and the suction header is closed (blue
star in Figure9) and when the power level is 20% of the nominal core power.
The former case has an important negative impact on the NC performance of the
RBMK; the latter case seems to represent the last possible NC operational point
for the RBMK, lower than the BWR corresponding point.(v)
When power level around 3% is considered, the two reactors behave in a
different way: it can be seen a lower mass flow rate in the RBMK enhancing a better
performance of the BWR with a low void fraction.


## 4.1. Application to the PSB-VVER Facility

As already showed by the previous examples (other can be got from, e.g., [8, 9]), the NCFM constitutes an advantageous tool to depict the NC
capability in removing the core power.  Starting from this point more applications have been done that cover
practically all the NPP types.In Figure5 (taken from [10]) it can be seen
the use of the NCFM for a VVER-1000 simulator whose scale factor is 1 : 300. The
experiment has been carried out in the frame of an OECD project in which the
core power and the SG pressure are maintained constant, while the mass is
stepwise drained. Both measured (in blue) and calculated date (in violet) are
reported. The experimental data goes out from the map because the flow rate
measurement systems were not qualified in two-phase flow giving a wrong
indication. However, the calculation confirms on one hand the suitability of
the NCFM and on the other hand the good NC performance of the VVER reactor
type.Figure 5
Natural circulation map for the PSB-VVER experimental and calculated data.


## 4.2. Revisiting the RD-14m Facility

In Section3, the RD14m data were considered but achieving results
far from what expected by the NCFM. Recently, a calculation has been repeated
fixing constant the core power and the SG pressure, while the primary side mass
is stepwise reduced, following a typical way to perform such kind of NC test.
The nodalization had been qualified against LBLOCA test (B-9401) carried out in
the framework of an IAEA project [11]. Analyzing the code results (summarized
in Figures 6 and 7), two main conclusions can be drawn.
(i)
The CANDU facility seems not so suitable for NC status (Figure6). However it should be
noted that the NCFM reasonably resumes the SG NC, it does not consider any
other mechanism simply because they are outside of its definition.(ii)
Figure7 shows
that in the CANDU installation other type of NC are present. The channel to
channel mode seems relevant from the core power removal point of view. This
behaviour explains why even at value of RM/V greater than 500 kg/m3,
where the NCFM predicts low flow condition (hence degraded capability in
removing the core power) the dryout is not experienced.Regarding the latter point (i.e., the channel to channel NC mode), Figure 7 reports the ratio channel mass flow rate over power versus time of two
different horizontal channels. It can be seen that once one core channel has
positive mass flow rate (i.e., the water follows the normal flow path entering
from the inlet and going out from the outlet of the channel), the other is
cooled by water circulating in the opposite path. This means that only a part
of the total core flow rate (sum of the core channels mass flow rate) reaches
the SG to be cooled.Figure 6
Natural circulation map for the RD-14m calculated data.Figure 7
Channel-to-channel natural circulation mode for the RD-14m calculated data.


## 4.3. Consideration of BWR Data

The last use of the NCFM regards the
consideration of boiling water reactor type coming from plant measurement (BWR)
and calculation results (RBMK).Figure8 reports the typical power-flow map of
a BWR in which the NC curve is marked. The maximum removable power in NC mode
is roughly 40% with a core flow rate of 25%.Figure 8
Typical BWR power-flow map.In Figure9 the BWR data are reported into the
NCFM (green dots and green square). Four points are present in the map: full
power condition (green square) at full pump speed, 25%, 4%, and 3% of the
nominal power (green dots). It can be seen that the points are practically
aligned along a line that marks the RM/V of the considered plant confirming the
suitability of the BWR to work in NC mode. Values of G/P comparable to what measured
in TP-flow conditions in PWR simulators are experienced in BWR but with a RM/V
ratio almost halved. The degradation of the performance in core removable power
typical of PWR is not present in BWR that in turns can be operated at 25% of its
nominal power.Figure 9
BWR and RBMK data and comparison with the NCFM.Again in Figure9 it can be seen also the RBMK data.  Those calculated values (red and blue stars are obtained by the use of a
qualified nodalization developed and tested during an EC funded project [12]. The
following consideration come out when calculated data are put inside the NCFM
and compared with BWR values.(i)
The RBMK data stay almost aligned at various power levels as occur in the BWR.(ii)
The full power point of both boiling water reactors (western and eastern
project type, resp.) has the (practically) same G/P ratio, but differs
by a factor of two in the RM/V ratio. This stresses the diverse project type
(vessel type versus channel type reactor), the different nominal point of work,
and the impact on the primary side (PS) mass of the two steam drums used in the
RBMK as separator.(iii)
The G/P ratio remains comparable between the two reactors at various power levels up to 4%.(iv)
Deviation from the consideration of the item above is experienced, when
the bypass between the pressure header and the suction header is closed (blue
star in Figure9) and when the power level is 20% of the nominal core power.
The former case has an important negative impact on the NC performance of the
RBMK; the latter case seems to represent the last possible NC operational point
for the RBMK, lower than the BWR corresponding point.(v)
When power level around 3% is considered, the two reactors behave in a
different way: it can be seen a lower mass flow rate in the RBMK enhancing a better
performance of the BWR with a low void fraction.


## 5. Conclusions

A system scaling study was the precursor of this work in which the
main objective was the setup of an engineering tool aimed at evaluating the NC
performance of an NPP. The NCFM was created gathering experimental data from six
PWR simulators.Recently, the use of the NCFM has been extended to practically all
types of reactor: VVER-1000 by the use of experimental data available after an
NC experiment carried out at the PSB-VVER test rig; CANDU revisiting the RD-14m
data after its nodalization qualification; BWR and RBMK comparing measured and
calculated data, respectively.Once more, the NCFM results in a very helpful tool to judge the NC
performance of different reactor types. It should be noted that the NCFM is
suitable to catch the SG NC type hence not able to deal with other NC type,
such as the channel-to-channel circulation experienced in the RD-14m.
Considering the characteristic and the way the NCFM has been derived, other applications
may regard support on scaling analysis, design and/or optimization of NC
system, and validation of computational tools.As mentioned above, this engineering tool has been directly derived
from the analysis of experimental data. However, a theoretical and generalized
relationship between the system inventory and the flow per unit power could be the
objective of a future research work.



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*Source: 479673-2008-04-20.xml*