Chapter
2 Systems of BWR Nuclear Power Plants
The reactor control system
of a BWR plant mainly consists of a reactor power control
system, a reactor pressure control system and a reactor water level
control system. Figure 2.6.1 shows a block diagram of these systems
including their subsystems. These systems, in concert with each
other, control reactor power level and keep reactor pressure and
reactor water level stable.
The reactor control system is designed in accordance with the
following basic design principles.
Reactor main parameters (power, pressure and water level) are
controlled to be within appropriate ranges during changes of plant
operating conditions and reactor load or other disturbances which
may occur during normal
operation.
Reactor power is controlled to meet required plant operating
conditions in light of the reactivity change brought by various
factors such as reactor load change, xenon concentration change,
high-to-low temperature change and fuel burnup.
Power oscillation is reliably and easily detected and accordingly
controlled.
Redundancy design (e.g.
triple-system) may be incorporated to enhance system reliability
and consequently plant availability for primary control systems
such as the recirculation flow control system, the reactor pressure
control system and the reactor water level control system since a
failure of a primary control system may cause a plant trip or power
decrease.
Reactor power control system
The reactor power control
system in a BWR consists of control rods and control rod
drive
*
SLRG:
Steam Line Resonance Compensator ** MSIV:
Main Steam Isolation Valve
Reactor
pressure
vessel
Steam
dryer
Reactor
Jet
pump
Steam-water
separator
Safety
relief valve
Reactor
M
Recirculation
!
pump J
Control I
rod
drive I
ij|
|
Control rod
M
Control
rod selection circuit
Control
rod drive system
Scram
signa), etc.
Recirculation
flow control system
Speed
controller
Hydr
aulic
coujil-
Scoop
tube position controller
Turbine
control system/ Reactor pressure control system
Pressure
deviation
regulator
SLRG*
Turbine
inlet
pressure
I
Bias
Load
deviation signal
L
MSIV**
Main
steam
flow
Feedwater
line I
set
point adjuster
Feedwater
hee
ler
Bypass
valve servo
Speed
demand
deviation
limi
ter
Feedforward
circuit
Turbine
steam
stop
L„_
Main
steam line
^Condenser
■Manual
Main
Reactor
water level
Tuitune
control
valve
l
Feedwater pump _
Controller
turbine controller
To
the
other
loop
Reactor
water control system
level signal ——
Feedwater
pump
Feedwater
pump
turbine
Figure
2.6.1 Reactor control systems
Main
controller
Speed
demanri
signal
limiter
Turbine
bypass
valvo
Control
valve
Speed
set point
Load
set point
Speed/load
signal
Feedwater
flow
L
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Instrumentation and Control System
Reactor Instrumentation and Control System
Plant control system
systems and a recirculation flow control system. The control rods
and control rod drive systems control reactor power level and power
distribution in the reactor core by adjusting the positions of the
control rods (which act as neutron absorber) in the reactor core. On
the other hand, the recirculation flow control system controls
reactor power level by changing recirculation flow rate and thereby
reactor core flow rate to resultingly control the average density of
the water (i.e. neutron moderator) in the reactor core. This
recirculation flow control system enables reactor power level to
rapidly change over a wide range of power levels almost without
altering the power distribution in the reactor core, which is one of
the major features of BWRs.
Figure 2.6.2 shows the relation between reactor core flow and
reactor power for typical cases of control rod operation and
recirculation flow change. When control rods are operated with
recirculation flow kept constant, reactor power changes along the
curves A, B or C in the figure;
and it changes nearly in proportion to core flow along the lines D,
E or F when recirculation flow is changed without control rod
position change.
Control rods are used for large power changes such as reactor
startup and shutdown, power distribution adjustment, or reactivity
compensation for fuel burnup; while recirculation flow control is
used for power to follow reactor load changes.
Control rods and control rod drive system
In a 1,100 MWe BWR plant,
185 control rods are installed in the reactor core. Each control rod
is provided with a control rod drive and a hydraulic control unit,
and is inserted or withdrawn from the bottom of the reactor core by
hydraulic pressure. Only one control rod can be operated at a time
during normal plant operation, and the rate of change of power level
by control rod operation is approximately 2%/min. All control rods
are simultaneously inserted into the reactor core promptly in the
event of an emergency shutdown of the reactor (i.e. a reactor
scram).
Reactor
power (%)
D,
E, F:
By recirculation flow control with constant control rod positions
A:
By control rod operation with constant recirculation pump speed
B:
By control rod operation at the minimum recirculation pump speed
C‘
By control rod operation at natural circulation with recirculation
pump stopped
Figure
2.6.2 Reactor power vs. core flow control curves
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In an ABWR plant, motor-driven fine motion control rod drives
(FMCRDs) are installed in lieu of the conventional control rod
drives. Rod positions are adjusted by insertion or withdrawal using
the FMCRDs, which provides finer rod position adjustment than the
conventional system does. During plant startup and shutdown, up to
26 control rods are operated at a time as gang mode operation, and
can also be operated automatically according to the predetermined
rod sequence. In the event of an emergency shutdown of the reactor,
all control rods are inserted promptly by the hydraulic control
units as in conventional BWRs.
Recirculation flow control system
Recirculation flow is controlled by two primary
loop recirculation pumps. As shown in Figure 2.6.1, the
recirculation flow control system for one loop consists of a
recirculation pump M-G set
comprising a variable frequency generator coupled with a
driving motor via a hydraulic coupling for torque control, and a
main controller and a speed controller for the M-G set. The
recirculation flow control system controls recirculation flow by
adjusting the speed of the recirculation pump through changing the
supplied power frequency of the induction motor that drives the
pump.
The power change demand signal is input to the main controller as a
load deviation signal from the reactor pressure control system or as
manual input. The main controller processes this signal by PID
(proportional integral differential) control and produces a speed
demand signal to the speed controller. The speed controller controls
transfer torque by adjusting the position of the scoop tube in the
hydraulic coupling so that the speed of the variable frequency
generator matches this speed demand signal. Thus, the frequency of
the variable frequency generator changes and hence the speed of the
recirculation pump changes.
Hie recirculation flow
control system is capable of changing reactor power level at a rate
of up to 30%power/min within the power range from 100% to about 65%.
In a plant where a static-type variable frequency power supply (VWF:
variable voltage variable frequency) system is installed as a power
supply
for the recirculation pumps instead of the recirculation pump M-G
sets, the output of the main controller is directly input to the VWF
controller as a speed demand signal. Thus, the output of the VWF
is adjusted and hence changes the recirculation pump speed.
In an ABWR, ten internal pumps are installed instead of the primary
loop recirculation pumps. Each internal pump is provided with one
VWF for power supply. As in conventional BWR plants, the power
change demand signal is input to the main controller as a load
deviation signal from the reactor pressure control system or as
manual input The output signal of the main controller is used as a
reactor core flow demand signal and compared with the feedback
signal of reactor core flow. Thus, VWF output is adjusted to control
the speed of the internal pumps so that actual reactor core flow is
adjusted to its demanded value. The internal pumps have smaller
inertia than conventional BWR pumps and hence better responsiveness
for power change, so the recirculation flow control system is
capable of changing reactor power level at a rate of up to
60%power/min. In addition, it is also capable of keeping 100% power
operation with operable nine internal pumps if one of the ten
internal pumps fails.
Reactor pressure control and turbine control system
Reactor pressure is automatically controlled to be constant in
combination with turbine control. As shown in Figure 2.6.1, the
pressure regulator compares the turbine inlet pressure with the
pressure set point and generates a pressure deviation signal.
Normally this pressure deviation signal is used to keep the turbine
inlet pressure constant by opening or closing the turbine
control valves (TCVs) and the turbine bypass valves
(TBVs), since in BWR plants pressure control has priority over
turbine control in light of reactivity control. However, in an event
such as a load rejection in
which turbine speed rapidly increases, the speed/load signal is used
in controlling the TCVs and the TBVs, taking precedence over the
pressure deviation signal through a low value gate.
A BWR plant with TBVs of 100% capacity of
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rated steam flow can handle a full load rejection upon a power grid
failure without a reactor scram by bypassing steam to the condensers
via the TBVs as the TCVs are fully closed, and transition to house
load operation in a controlled fashion. This ensures that the plant
is quickly returned to power operation once the failure in the power
grid is corrected.
In an ABWR plant, the pressure
regulator utilizes reactor dome pressure for comparison with the
pressure set point instead of the turbine inlet pressure which is
used in a conventional BWR plant. The dome pressure control enables
more stable reactor pressure control in the ABWR than in the BWR
Reactor water level control system
The reactor water level
control system controls reactor water level to be
constant within the acceptable range to keep the efficiency of steam
water separators by controlling reactor feedwater flow with a
three-element control
method utilizing a reactor water level signal, a reactor
feedwater flow signal and a main steam flow signal. This method
provides stable and quick control of reactor water level on the
basis that reactor water level change is estimated from the detected
mismatch of feedwater flow
and main steam flow and consequently the difference between the
flows going into and out of the reactor. At a lower reactor power
level, water level is controlled with a single
mode control method utilizing only a water level signal.
Instrumentation and
control system for the reactor safety protection system (RPS) The
RPS initiates appropriate safety and protection operations to
prevent or mitigate the adverse conditions that could affect
reactor safety when such abnormal transients or malfunctions occur
or are anticipated to occur. The RPS consists of a reactor
emergency shutdown
(reactor scram) system and an initiation system for the
engineered safety features (ESFs) such as the emergency core
cooling system (ECCS) (Refer to sections 2.7.1 and 2.7.2).
The RPS is designed in accordance with the following basic design
principles.
(a)When abnormal transients
occur during
plant operation, the RPS is capable of sensing them and
automatically initiating the reactor emergency shutdown system to
ensure that the specified acceptable fuel design limits are not
exceeded.
The RPS is designed to ensure that the specified acceptable fuel
design limits are not exceeded in the event of any single
malfunction of the reactor emergency shutdown system such as
accidental withdrawal of the control rods.
Under accident conditions, the RPS promptly senses the abnormal
situation and automatically initiates the reactor emergency
shutdown system and the ESFs.
The design of the RPS incorporates sufficient redundancy and
electrical and physical independency so that no possible single
failure or off-line state of any single component or channel in the
system results in hindering the safety and protection functions.
The RPS is designed to sustain an acceptable safe state (fail-safe
or fail-as-is) if it is disconnected or loses power.
©Where practicable, the
RPS is separated from non-safety-related instrumentation and control
systems. If some part is in common, the RPS is not affected by a
failure of any non-safety- related instrumentation and control
system.
The RPS is designed to provide periodical testability for its
functions during normal operation.
Seismic design is incorporated in the RPS design.
Reactor shutdown system (or reactor emergency shutdown system)
The reactor (emergency)
shutdown system consists of two channels (channels A and
B) as shown in Figure 2.6.3. Each channel has at least two
independent tripping contact points for a single measured parameter.
Actuation of either tripping contact point hips
the applicable channel, and a simultaneous trip of the two channels
leads to a reactor scram (i.e. one-out-of-two-twice logic).
The reactor scram mechanism is as follows. A reactor scram signal
actuates scram valves to rapidly insert control rods. The scram
valve is actuated by a scram pilot valve with two
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Chapter
2 Systems of BWR Nuclear Power Plants
solenoids. When one or both solenoids are energized, this pilot
valve keeps the scram valve at the closed position with air pressure
applied on the diaphragm of the scram valve. When both solenoids are
de-energized, the air pressure is released from the diaphragm to
open the scram valve and consequently the control rod is inserted
immediately. A simultaneous trip of the two channels of the reactor
shutdown system de-energizes the two solenoids of the pilot valve
for each control rod and thus leads to a reactor scram. Meanwhile, a
single channel trip deenergizes only one solenoid and hence
does not cause a reactor scram. Table 2.6.1 lists the trip signals
of the reactor shutdown system.
In an ABWR plant, microprocessor-based digital systems are employed
for the trip channels and logic
circuits of the RPS instead of the conventional relay sequences and
analog circuits. As shown in Figure 2.6.4, the trip channels are
completely separated into four independent divisions, and
two-out-of-four logic is implemented instead of one-out-of-two-twice
logic. Such digital systems along with self-diagnostic functions are
expected to enhance system reliability.
In implementing software logic in the safety protection system,
software production is noted and accordingly a particular method is
employed for verification and validation. The
method is referred to as V & V (Verification & Validation),
which is detailed in JEAG4609 (1999) "Guide for Application of
Digital Computers in the Safety Protection System” of the
Technical Guidelines of the Japan Electric Association.
Figure
2.6.3 Reactor shutdown system trip channels
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Table
2.6.1 Reactor shutdown system trip signals |
Detector |
Purpose |
Reactor high pressure |
Pressure switch |
To detect a rise in reactor pressure beyond the normal pressure control range caused by abnormal conditions such as unexpected isolation or power increase of the reactor in order to prevent the reactor from further developing abnormal conditions |
Reactor low water level |
Level switch |
To detect reactor water level lower than the normal control range caused by abnormal conditions such as reactor water level control failure or reactor coolant leakage in order to secure adequate reactor core cooling |
Drywell high pressure |
Pressure switch |
To promptly confirm a leakage occurrence by detecting a rise in drywell pressure resulted from reactor coolant leakage from the reactor coolant pressure boundary in order for measures to be taken such as leakage expansion prevention or coolant replenishment |
High neutron flux |
APRM IRM (note 2) |
To prevent expansion of abnormal conditions by detecting a rise in neutron flux caused by abnormal reactor conditions |
High neutron flux (note 1) |
||
Neutron monitoring inoperable |
APRM IRM (note 2) |
To shut down the reactor for safety before operation continues without power monitoring of the reactor core |
Scram discharge volume high water level |
Level switch |
To shut down the reactor for safety before operation continues with potential incapability to scram |
Main steam isolation valve (MSIV) closure |
Valve position switch |
To control power increase caused by a reactor pressure rise with the valves closed |
Turbine stop valve closure |
Valve position switch |
|
Turbine control valve fast closure |
Pressure switch Position switch |
|
Main steam line high radiation |
Gamma ray detector |
To control transfer to the turbine of the radioactive materials brought about by fuel damage |
High seismic acceleration |
Acceleration sensor |
To shut down the reactor for safety against potential equipment failures caused by an earthquake |
Manual scram |
Pushbutton switch |
To shut down the reactor by operator judgment |
Reactor mode switch at shutdown | |
Reactor mode switch |
To keep the reactor in the shutdown mode |
(Note
1) "SRNM short
period1'
for plants with SRNM installed
(Note 2) "SRNM1’
for plants with SRNM installed
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Figure
2.6.4 Reactor shutdown system trip channels (ABWR)
ESF initiation system
ESF actuation structure
The ESF initiation system
initiates appropriate safety and protection operations of
the ESFs to prevent or mitigate the adverse conditions that could
affect reactor safety when such abnormal transients or malfunctions
occur or are anticipated to occur.
hlie ESFs initiation system
consists of two or of four channels of analog modules and two
channels of logic circuits. Each channel is physically and
electrically separated from the other channels. As shown in Figure
2.6.5, appropriate systems and/ or functions in the ESFs are
initiated according to the output of the logic circuits. For
example, the high pressure core spray system (HPCS) is initiated by
the “reactor vessel low water
level” signal or "drywell
high pressure” signal. The “reactor vessel low water level”
signal is generated from four water level analog detectors and trip
circuits through the one-out-of two-twice logic. Similarly the
“drywell high pressure” signal is from four pressure detectors
and trip circuits,
ESF initiation signals and functions
The ESFs are initiated
basically upon detection of an abnormally low reactor water level or
an abnormally high drywell pressure as shown in Figure 2.6.5. Such
abnormal situations may be brought about in a loss
of coolant accident (LOCA) resulted from a reactor
primary system breakage or the like. The low pressure core spray
system (LPCS) and the low pressure core injection system (LPCI, one
of the operation modes of the residual heat removal system
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(HPDG:
HPCS diesel generator)
|~TD1
OR
Time
delay
Figure
2.6.5 ESF initiation system
(RHR)) start coolant injection to the reactor core when reactor
pressure becomes as low as the discharge pressure of the pumps
of these systems. Hie
automatic depressurization system (ADS) has a function to
depressurize the reactor and it is initiated according to the
output of logical operation AND of the three signals: “reactor
vessel low water level”, “drywell high pressure” and “one
of the low pressure injection systems operating”.
In addition, the ESFs include the following functions:
to isolate the reactor by closing isolation valves other than
the main steam isolation valves (MSIVs);
to start up the standby gas treatment system
(SGTS) and keep the reactor building at negative pressure so as
to prevent radioactive material leakage to the outside; and
to start up the HPCS diesel generator and the emergency diesel
generators to secure emergency power supply.
Table 2.6.2 summarizes the ESF initiation signals and relevant
protection functions.
In an ABWR plant, the ESF initiation system consists of
microprocessor-based digital systems
and employs the two-out-of-four logic as is the case with the
reactor emergency shutdown system.
Table
2.6.2 ESF initiation signals |
Protection Function |
D3 L-2 Reactor low water level Lrl |
SGTS initiation Isolation valves closure (except for MSIV) |
MSIV closure HPCS and HPCS diesel generator initiation |
|
LPCS initiation LPCI initiation ADS actuation Emergency diesel generator initiation |
|
Drywell high pressure |
HPCS and HPCS diesel generator initiation LPCS initiation LPCI initiation ADS actuation SGTS initiation Emergency diesel generator initiation Isolation valves closure (except for MSIV) |
Main steam line low pressure |
MSIV closure |
Main steam line high flow |
MSIV closure |
Condenser low vacuum |
MSIV closure |
Main steam line high radiation |
MSIV closure |
Main steam line high tunnel temperature |
MSIV closure |
Reactor building high radiation |
SGTS initiation |
