International Conference on Nuclear Engineering Volume 2 ASME 1996
Control-rod,
Pressure
Analyses
and Flow-Induced
of a Direct-Cycle,
Light-Water-Cooled
KAZUAKI
KITOH,
Accident
Supercritlcal-Pressure, Fast Breeder
SEIICHI
and Transient
and YOSHIAKI
KOSHIZUKA
Nuclear Engineering
Reactor
OKA
Research Laboratory
The University of Tokyo Tokai-mura, lbaraki JAPAN
ABSTRACT
is used for the power transients
The features
of the direct-cycle,
supercritical-pressure.
light-water-cooled
fast breeder reactor
thermal
and simple
efficiency
principle
is basically reactor.
basic
requirement
safety
“Maintaining
rod, pressure
analyzed
though
pressure
in this paper.
system
partial
is also
analyzed as the transients.
The behavior
constant
pressure.
accident.
density
breeding
valves). control-
and the
all the transients
and
reac-
simple reactor system
The core can be designed
is identical
can be maximized
and
as thermal
supercritical
water does not exhibit
recirculation
system, steam separator.
reactor pressure vessel and control
time
of a pressurized similar
of the SCFBR is
water reactor
number of coolant
537
a change
fossil-fired
them. The The
of phase. The
and dryer of a boiling Roughly
speaking.
rods are similar
power
the
to those
(PWR). the containment
to a BWR. and the balance of plant
supercritical-pressure
flow
among
m the fast converter.
water reactor (BWR) are unnecessary.
loss of flow
less than one tenth of a BWR in which the recirculation
capability”.
power output
power are
the coast-down
of the supercritical-water-cooled
(SCFR). The plant system
feedwater pump,
from the analyses
coefficient
17Ocallg
and lower
reactor3 (SCLWR). fast breeder’ (SCFBR) and fast converter’
loss of feedwater
10s and to cope with the total The coolant
Loss of load
Fly wheels should be equipped
with the feedwater pumps to prolong more than
the safety criteria.
tors are high thermal efficiency,
of the flow-induced
is not so much different
accidents
The advantages
Total loss of flow
flow and loss of external
below
1.0. In conclusion,
is 104.9cal/g
INTRODUCTION
of the rapid reactivity-induced
start of an auxiliary
fuel enthalpy
MDHFR above satisfy
the
inducing control
loss of coolant
assuming
MDHFR is 1.66 which are sufficiently
distribution
inadvertent
transients
The maximum fuel enthalpy
than the criteria:
pressure
from the normal operation
the smallest
of fuel temperature
turbine by-pass
transient
are
1.0 and
abnormalities,
is analyzed.
the turbine
from the normal operation.
rod withdrawal
maximum
the reactivity
are
using
(with and without opening
ratio below 1.10 of 275MPa. Among
pressure
MDHFR above
at
at ICONE-3.
and pump seizure are analyzed as the accidents.
heating,
control
(MDHFR) and the maximum
lower than the criteria;
for operation.
The computer code
such as control rod withdrawal.
rod withdrawal
pressure
The results of flow-
to handle the pressure change
regulation
for the analysis
transients
behaviors
abnormalities
change was neglected.
valve. The change considered
type. The transient
of SCFBR were reported
has been improved the pressure
sufficiently
since its coolant
and flow-induced
and presented
induced transients
The safety
the core flow” is the
of the reactor.
system is the one through control
heat flux ratio
the same as that of an LWR since it is a
water-cooled
The over pressurization
BWR. The power reaches 120%. The minimum deterioration
(SCFBR) are high
reactor system.
control.
at the loss of load is not so severe as that of a
is
is similar
to a
plant VW.
The
tines is only two because
of the high
coolant
enthalpy.
Containment
The thermal efficiency The coolant
void reactivity
zirconium-hydride
volume is much reduced.
is improved
coolant
by 24% over a BWR’.
is negative
by placing
thin
layers between seeds and blanket’?
fundamental
fast breeder reactor
requirement
for safety
The
transients.
be different
generated
of control
rod. pressure
of the direct-cycle
breeder
characteristics analyses
reactor
fuel temperature
are summarized
and
The
in Table
SCFBR
satisfied
core
the computer
the heat
temperature
rises
Since the SCFBR has no recirculation
line
loop. the coolant
flow in the core should be
If the following
under any condition.
two requirements
the excessive
are
increase
of
(1) To keep the feedwater flow from the coldleg (2) To keep the coolant outlet open at the hotleg
are not Abnormal
code has
been improved to handle the pressure change inducing system using the turbine control
and the change of fuel temperature
by removing
1. Preliminary
These assumptions
Therefore
the
of anticipated
the cladding surface temperature is avoided:
pressure and steady state
distribution’.
pressure regulation
This is accomplished
directly maintained.
events of the SCFBR were
constant
conservative.
accidents
supercritical-water-cooled
(SCFBR).
of the flow-induced
reported by assuming
always
and flow-induced
is kept by
to maintain
under any condition
in the core before the cladding
nor primary
This paper deals with the safety design and analyses
fast
integrity
excessively.
from the current LWR’s.
transients
pressure
The safety system should be designed fuel cladding
(LMFBR). The
should
The operating
means of turbine control valve.
power costs will be much reduced compared with those of a LWR and a liquid-metal
to the core at 25.OMPa. The whole outlet coolant
flows to the turbine.
the
events are classified
their frequencies
valve
are relatively
when their frequencies
distribution.
designed
into transients
high,
are low.
Auxiliary
to keep the safety when
when
or into accidents systems
abnormal
are
transients
occur. As far as the plant safety is kept by the auxiliary Table1 Characteristics core
of SCFBR
Coolant inlet/outlet temperature(%) Coolant inlet/outlet density(g/cn?) ~~~\~2~.~,23,3 Coolant inlet/outlet velocfty(m/s) MoxlQ.00ll .Ol Fuel/rod diameter/pitch(cm) SSlO.052 Claddinq/thickness(cm) Plutonium fissile enrichment, inner/outer seed(%) 14.9606.52 Average discharge bumup(GWd/t) 77.7 Coolant density coefficient BOIEC/EOIEC( Aldk - (s/cmj’) 0.0526/0.032 Doppler coefficient BOIEC/EOIEC(AkIkPC) -2.5x1rj5/-2.0x1d5 Maximum linear power(W/cm) 400 Power density maximum/average (W/cm? 453/t 72 Fuel centerline temperature(r) Maximum 1995 system T4 Thermal efficiency’ . . 0.415 Pressure(MPa) 25.0 Main steam line number 2 Feedwater flow rate(kg/s) 2646 BOIEC:Begining of initial equilibrium cycle EOIEC:End of initial equilibrium cycle
systems,
the plant
operation.
When
systems
Concept
SYSTEM
AND SAFETY
of SCFBR
safetv
abnormality
systems
has 50% capacity
are actuated
according
to its
of auxiliary
feedwater
system
of automatic
depressurization
level.
Flow level 2 : actuation (AFS). Flow level 3 : actuation
system (ADS) and the low pressure coolant injection
system (LPCI).
To keep the coolant outlet open, measured and the following to the abnormality
Pressure level I
the core pressure
valves are operated
is
according
level.
turbine control valve,
Pressure level 2 : turbine bypass valves. safety relief valves
DESIGN
depressurization
system
Turbine bypass
of two lines. each of which
control
valves
valve
(SRV) or automatic system
are used when
the reactor
control
Turbine valve
while ADS is
operated ICthe pressure decreases below 24.OMPa.
538
are
due to the loss of load. SRVs are open
if the reactor pressure rises above 265MPa.
and supply
(ADS).
24.OMPa and 26.5MPa.
are open if the turbine
closed: for instance,
of the steady state mass flow rate. The
main feedwater pumps are driven by turbines
safety
Flow level 1 : reactor scram,
The plant system of SCFBR is depicted in Fig 1. The consists
to the normal
engineered
To keep the feedwater. the flow rate is measured and the following
pressure stays between main coolant system
return
occur,
are actuated.
Pressure level 3 PLANT
can quickly accidents
One of the advantages that the detection the
basic
opening
of the present safety concept
of abnormality
requirements.
The inlet
of outlet are directly
with the measurements respectively.
is straightforward feedwater
recognized
and the
turbines
of the flow rate and the pressure,
misoperation
and
probabilistic
misunderstanding
though
relatively
quick
safety assessment
is
falls
and
needs active
actuation.
the level
The
accident
(LGCA)*. LPCI has four lines, each of which has 805 kg/s. In the LOCA analysis,
actuated.
the AFS capability.
used, ADS is actuated simultaneously.
When the reactor
feedwater or the incident the ADS+LPCI system
pressure
the events
is
operation
near the critical is markedly
is not kept above
pressure
valves are open, the coolant downcomer
cooling
The core flow is maintained
are managed
to satisfy
the safety
the following
conditions
of
of the reactor scram:
(1) mass flow rate at cold leg below 90% of the steady state,
the depressurization
(2) reactor power over 120% of the steady state.
by LPCI after the
reactor pressure is reduced to around atmospheric
the actuation
analyzed
are employed:
-Actuation
(3) loss of external power,
pressure.
(4) rapid closure of the turbine control valve.
Once the reactor pressure falls down, it needs a long time to return to the normal operation.
LPCI is
where the
When the relief
and flows through the
during
Actually.
in any case in the present study. since all
In the present analyses, actuation
stored in the lower plenum,
and coldlegs evaporates
core. This ensures
(22.lMPa)
deteriorated.
to the
criteria without ADS+LPCI.
24.OMPa. ADS+LPCI should be used as well to avoid the
heat transfer
two of four lines
break line and the other fails to start up. When LPCI is
is remained for the future
If AFS fails to keep enough
of LPCI was
of large break loss-of-coolant
capacity
not considered
is beyond
when the flow rate
3. The capacity
are assumed to lose the function; one is connected
study.
period.
below
Each
mass flow
is actuated when the flow rate is lower
determined by the analysis
both mechanical
it always
of the steady-state
than the level 2. LPCI is in operation
This straightforward
system will enhance
reliability,
components
of
in the control-room.
and simple safety and human
risks
svstem
of four lines; two of them are driven by
10% capacity
rate. This system
side and the
outlet is always at the hotleg side. This simple relation reduce
safetv
and the others are driven by electric motors.
line possesses
Besides, the flow path in the core is unique.
to
of SCFBR
AFS consists
by the operators
since the feedwater is always at the coldleg
good
Desian
is
with
Thus the probability
of
of ADS+LPCI should be small enough
to
(5) reactor period below 1Osec.
The scram including
enhance the load factor.
of the control
Control rod f-w
1
zif$’
is assumed
to be completed
the delay time of signal
relief
processing
rods. The scram reactivity
in
3.7~~
and motion
is $12.7.
Turbine by-pass \mlve Turbine control valve 1
I I
High pressure auxiliary
Main feedw a er pump Fig.1 Plant system of SCFBR
539
-Actuation
of the auxiliary
feedwater system
CALCULATION
(AFS):
MODEL
(I) mass flow rate at cold leg below 20% of the steady state,
The calculation
(2) trip of the main feedwater pumps.
following
It is assumed that two of four AFS lines are actuated
code is developed
control
and inertia of the coolant.
between 24.OMPa and 26.5MPa.
valve
within
the
pressure
(2) The heat transfer coefficient of the turbine bypass
valves.
the coolant
(1) rapid closure of the turbine control valve.
time of signal
processing.
of the turbine bypass
those of the turbine control
valves
perturbation
from cladding
is calculated
(3) The reactor
control
the
by
surface to
Dittus-Boelter
corre-
lation.
Turbine bypass valves are opened after 0. lsec because of the delay
on
(1) The reactor pressure is kept at 25.OMPa by the turbine
after 5.0sec because of the delay time of signal processing
-Actuation
based
assumptions.
Capacity
power
is calculated
by point
equation with six delayed neutron groups,
and
decay heat is calculated
are the same with
using
kinetics while the
two group
approxi-
mation of ANS+20% formula.
valve.
(4) The axial
reactor
power distribution
is assumed
to
follow the cosine distribution. -Actuation
of the safety relief valves (SRV):
(5) Doppler and coolant
density feedback are considered.
(1) setting
of pressure values are shown in Table 2.
(6) The reactor system
is divided into three parts: core.
Table2 Set point and number of safety relief valves
(7) The hottest single channel is analyzed.
upper and lower plenums, as shown in Fig. 2.
close(MPa)
opan(MPa)
number
26.5
25.5
2
26.7
25.7
3
26.9
25.9
10
27.1
26.1
10
The core is expressed
includes
down comer is modeled
large volume.
equilibrium
cycle
equilibrium
cycle
(BOIEC)
and
the
BOIEC the fuel temperature
coefficient
coolant
is larger.
density
end
(EOIEC) are analyzed.
coefficient
of
is smaller
It is impossible
In the present
are the inlet coolant
temperature,
of initial
Since
by a single
cell. The
to calculate
the density
The heat transfer
from Dittus-Boelter
initial
The flowchart
and the
Fig.3.
more severe results
pressure.
calculation.
flow and the inlet coolant coefficient
at pseudocritical
of the calculation
The calculation
reactor
are expected at the BOIEC.
the calculation
is calculated
formula which gives a conservative
heat transfer coefficient
at the
analysis,
from the core inlet to the outlet. The boundary
conditions pressure and flow-
at the beginning
mode1 and
change in the upper plenum with a single cell, since it has
proceeds
induced accidents and transients
channel
The lower plenum which
upper plenum which includes main steam lines is divided into five nodes.
In the present study. the following
by single
divided into five nodes axially.
consists
reactivity
temperature. code is shown
of the thermal feedback
and
in
hydraulic. nuclear
The input data are the inlet coolant
flow and
the inlet coolant temperature. The analyzed accidents are as follows: (1) Total loss of reactor coolant
flow;
Thermal
(2) Reactor coolant pump shaft seizure; The analyzed transients
and reactor
oressure
cal-
and mass conservation
equations
are as
are:
(3) loss of feedwater heating; (4) inadvertent
hvdraulic
culation
The energy
start-up of auxiliary
(5) partial loss of reactor coolant
feedwater system:
follows: d
flow;
u-(@/)=Qour-V(W
H)
(6) loss of external power: (7) loss of load (turbine bypass
“L, dr
valves are opened):
w
(8) loss of load (turbine bypass valves can not be opened);
where,
(9) control rod withdrawal
H
(from normal operation):
540
: coolant enthalpy(J/kg).
(1)
(2)
Qout t
: heat transfer from fuel surface to coolant(W). : time(sec),
V
: cell volume(m).
W
: mass flow rate(kg/s).
p
: coolant density(kg/m?.
The coolant calculated
enthalpy
and mass flow rate in the nodes are
from the above equations.
The opening calculated
ratio
of turbine
from the following
control
valve
is
equations:
V=G(s).Vr.
(3)
G(s) = s
where. G(s) : transfer function. V : angle
that
between Lower plenum
is defined as valve
the present
capacity
control valve(%).
I
Vr : required angle of turbine control valve(%)
Fig.2 Calculation model
The change of pellet average temperature from the following c START
V
~Twe
)
Hydraulic
=
is calculated
equation:
Qpeller - Qour CD.0 Y
(5)
where,
Initial Condition
Thermal
ratio
and the full one. of turbine
Cp
Calculation
: heat capacity(J/kg”C),
Qpellet
: generated heat in the pellet(W),
Tave
: pellet average temperature(c).
The pellet related
centerline
and surface
to the pellet average
temperatures
temperature
are
and the heat
transfer to the coolant as follows: Tcenter = Tave + $
? j”
Rpr;cp
- v
,
r;q”’ Tsurfoce = Tcenrer - 4~.
8K,
(6)
ATme.
npr’cp + I 4K, ATwe.
where, I?,
: average thermal conductivity
6’
: power density(WJm3.
rr
: fuel radius(m),
Tcenter Tsurface
Lp-Calculate
Value of Turbine
Control
: pellet centerline
of fuel(W/mc).
temperatureK)
pellet surface temperature(c).
Valves The cladding surface temperature is calculated from the following
equation.
(8) where.
( Fig.3 Calculation flowchart
541
D
: fuel rod diameter (m).
h,
: heat transfer coefficient
(W/m”C) I
Tclad
cladding surface temperature
Tcoolant
coolant
temperature (“c).
(“c)
The heat transfer coefficient
is calculated from Dittus-
Boelter formula.
does not occur under the postulated
normal Reactivitv
feedback
The Doppler
calculation
feedback
and coolant
condition
under the postulated
safety criteria are made in referring
density
reactivities are calculated from the following k+*, = a~,,(Tove.~,Q). Alive. k*urn
feedback
(1) Stainless (9)
=c&,@).Aa,
transients.
The
to those of the LWR:
(10)
Steel (SS) cladding surface temperature
below
1260°C. (2) Reactor pressure below 110% of 27.5MPa.
maximum
pressure for operation.
6
: average
coolant
density
of all
weighted by the power distribution AE
: difference
between
coolant density k-
five
cells
(kg/m?,
the average
Transjents
(1) Minimum deterioration
and initial
(kg/m?.
Tave
(2) Reactor pressure below 105% of 27.5MPa.
density feedback,
: average
fuel temperature
temperature CJ-
:
of all
five
cells
(3) Maximum
(c).
fuel enthalpy
The stainless
(“c) , that
gives
fuel
that
gives
coolant
temperature
from
density feedback
as
fo1lows:
deterioration
(11)
where, keff : effective multiplication
phenomena
do not exist.
factor.
At the supercritical
continuously
and boiling
The specific heat shows a sharp temperature.
The heat transfer
nearby
this
temperature
increases
locally
and continuously
occurs.
temperature.
This behavior
The wall where the
is much milder than
The MDHFR is used in a similar way like the
minimum critical heat flux ratio (MCHFR) for the BWR. In this study, Yamagata’s
Core power is calculated from the point kinetic
equa-
of the ANS+20% evaluation
as shown
in the following
(12)
{i=1.2}.
correlation
where,
(13)
G
: mass velocity
Q
: critical heat flux (kW/m$
(kg/mh) ,
The reasons for choosing ratio
of group
i of decay
the limit MDHFR= 1.OO are:
heat
(constant),
(a) The cladding
D, : power of group i of decay heat (W),
temperature
to the continuous
: reactor power (W).
supercrltical
of group i of decay heat (/s).
(b) Yamagata’s channel
CRITERIA
The safety philosophy
is used to evaluate the
heat flux?
where.
equation. $D, =k,(a;Q-D,)
deterioration 4, = o.2ciz.
tion. Decay heat is calculated by two group approximation
h , : time constant
heat flux where the
occurs
deterioration
calculation
power
is
commission’.
deterioration
the burnout.
SAFETY
occurs.
pressure, water density changes
peak at the pseudocritical + I .O,
criterion
The MDHFR is defined as the ratio of the cladding
density
factor is calculated
feedback and the coolant
: initial
X
for the LWRs with SS
surface heat flux to the deterioration
multiplication
keff = km
