Chapter
5 Operation and Maintenance of PWR Plants
to the synchronous speed and the generator is synchronized with the
power transmission grid ("generator tie-in” operation).
During the turbinegenerator startup operations, excess steam is
bypassed to the condenser via the turbine bypass line. When the
generator output exceeded the level of station auxiliary loads,
power supply to the auxiliary loads is switched over from the
offsite power supply (startup transformer) line to the
generator (the station auxiliary transformer) line. Typical startup
curves of a PWR plant are
shown in Figure5.1.2.
These are examples where the power increase rate is not restricted
to ensure the fuel mechanical integrity. In cases of reactor
startups after refueling,
or power increases after long periods of low power operation, the
power increase rate is restricted so as not to exceed 3%/h during
the period of reactor power increase from 20%
to 100%. However, if a plant has been operated at a specified power
level for more than 72 hours, the restriction is ignored below that
power level.
In actual plant operations, the above-mentioned restriction is
applied more conservatively, although practices vary plant by plant
Figure 5.1.3 shows an example set of startup curves of plants under
the restriction.
The main duty of the operating crew during the plant normal
operation is to compensate the reactivity loss of core which
increases as fuel burnup proceeds. The reactivity of a PWR plant
reactor is controlled by two methods: operation of control rod
clusters and adjustment of boron concentration in the reactor
coolant. These two methods are used for different purposes. The
control rod clusters are used for control of the core reactivity
during relatively rapid reactivity changes as required in startup,
shutdown or emergency shutdown phases of reactor operation. They are
also used for the control of core reactivity during rapid load
changes together with the reactor control system. The adjustment of
boron concentration is used for the control of core reactivity
during slow reactivity changes caused by the fuel burnup, the
buildup or decay of fission products which have relatively large
neutron absorption cross sections, such as xenon and samarium, the
reactivity changes
during the reactor coolant temperature change from room temperature
to the zero-power operation temperature due to the moderator
temperature coefficient, and so on.
The control rod clusters are classified into two groups, a shutdown
group and a control group, according to their purposes. Control rods
of the shutdown group are used only for the emergency shutdown of
the reactor, and hence, they are fully withdrawn during the plant
power operation. The control rods of the control group are
automatically controlled by control rod control system signals
during the plant normal operation. Both groups of control rod
clusters are divided into several subgroups, called banks, for the
purpose of restricting reactivity added to the core at one time.
Withdrawal operation procedures of each bank of the control group
rod clusters are fixed and are not changed throughout the core
lifetime. Since all control rod clusters are almost fully withdrawn
during the plant normal operation and subsequently the reactor core
neutron flux distribution pattern is flat, flux distribution does
not need to be adjusted in anyway
including computer control.
Adjustment of the reactor coolant boron concentration is conducted
by batch operations in the chemical and volume control system, i.e.
adding a volume of boric acid solution to increase the concentration
and adding a volume of pure water to decrease it
Plant normal operation usually means that the plant is steadily
operated with its rated power, except during periodical intermittent
power reductions for the turbine stem-free testing operations. In
plant normal operation, the main operational procedures which the
crew must implement are the dilution and concentration of reactor
coolant boron to compensate for core reactivity changes due to fuel
burnup and xenon buildup or depletion following reactor power
changes. For instance, when reactor core reactivity gradually
declines as the core fuel burns up, the average temperature of the
reactor coolant gradually decreases, and the control group rod
clusters are automatically withdrawn to compensate for this. When
control rod clusters are withdrawn and reach certain specified
positions, plant operators are required to dilute the boron
concentration in the reactor coolant and restore the
5
—
5
NSRA,
Japan
Plant Normal Operation
(Shujui)
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Pressurizer Water Level
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Achievement
of Rated Power
— Automatic
Feed Water Control, Automatic Control Rod Control
Turbine Bypass Control
Change over toTAVC
Control
Tie-in, Low Genereator Load
Turbine Speed up Initiation
—* Rated R.P.M
Criticality, Specified
Power Level Maintain by Control Rods
and Boron Concentration
Steam Supply Initiation
(Condenser Vacuum Establish)
Main Steam Pipe Heat up
Zero Power Coolant Average
Temperature and Pressure Reached
— Turbine
Bypass Control Change over to Pressure Control
Isolation of Residual Heat
Removal System
Charging Flow Automatic
Control Normal Letdown Line
Alignment Establish
Pressurizer Water Level Set
at Zero Power Level
Pressurizer Vapor Formation
Initiation
Figure
5.1.2 PWR standard startup curve
Residual Heat Removal Pump
Shutdown
Hydrogen Blanket Formation
in tire Volume Control
Tank
Primary Coolant Water
Quality Adjustment Complete
Primary Coolant Water
Quality Adjustment
Initiation (Oxygen Concentration
etc.)
Primary Coolant Pump Startup
All Pressurizer Heaters
Puton Control
Rod
Shutdown Bank
Withdrawal
Charging Pump Startup
Let Down Line Pressure
Setpoint Adjustment
I
L i |
8 |
(%) pAOq JozrJnsso |
8 i |
B CM . 1 |
(a) ajnpHodmaxSQH |
8 |
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G.UI3/3T arnssay gOH |
System Checkup prior to Plant
Startup Complete
NSRA,
Japan
5~6
§
3
100
(Startup
Time) 1.
At cold shutdown: (Suberitical
with coolant temperature of 60°F
or under)
2.
At hot shut down: (Subcritical
with coolant temperature of 286“C)
NSRA,
Japan
286t
50f
300
200k
RCS
Cool down
0
100
Actual
Time
Required
Time |
Required Time |
Duration Time |
|||
® Heat up Initiation of heat up and pressure raise ~ Rated temperature and pressure |
30 h |
30 h |
|||
® llcat up complete ~ Criticality |
0 |
39 |
|||
CO Criticality — Tic in |
3 |
42 |
|||
® Tie-in — Full Power |
(1) Power raise lime |
150 |
200-202 |
||
(2) Water quality adjusting time (al 30% power) |
8 ~ 10 |
||||
Procedure |
Required Time |
Duration Time |
|||
(D Critical procedure initiation — Critical!ty |
0.5 h |
0.5 h |
|||
(g) Critiacal i ty ~ Tie-in |
1.0 |
1.5 |
|||
® Tie-in ~ Full Power |
34 |
36 |
|||
100$
Initiation
of
Coo!
down Complete
35
-1U--5-
7h
8h
(Power
Change Rate etc.)
7 5$
50-60
r
50-60
TC
i ?i r
50
1.
Startup from Long Time (Over 2 weeks) Shutdown.
Power
Increase Rule
shall
be
Hold
75%
Electric Power over
Power
Increase Rale
shall be
RCS
Water qua!ity
check
[leal
up Initiation
80
r
Turbine
RPH
increase
Turbine
1,800
rpm
Under
10%
: Control
rod bank D shall be withdrawn over
80% during this period.
=
§
as
286
r I
. *20
3
%/h
from Tie-in to 75% Electric
Power.
75h
Period of Time.
0.5
%/h
from 75%
Electric Power or Over.
2.
Startup from Short Time (Under 2 weeks) Shutdown.
®
Power Increase Rale shall
be 3%
from Tie-in to Tull
Power.
1
Power Decrease Rate shall
be 10 %/h.
30h
9h
3$/h
0.5$/h
3h -
*
These figures shall be changed depend on the reactor operation time
and shutdown
conditions.
Figure
5.1.3 Startup and shutdown curve under restricted power raise
conditions
Chapter
5
O
S
I
§
positions of control rod clusters. Through repeating such
operations, the boron concentration in the reactor coolant linearly
decreases as the reactor core approaches the end of its service
life.
The responses of a reactor plant in the case of sudden off-site
power rejection due to a system disturbance or some other occurrence
are briefly discussed below. When an off-site power rejection occurs
two things take place: a power deviation signal based on the
discrepancy between the turbine output and the reactor output
generates a signal to insert rod clusters at the maximum speed to
reduce the reactor output, and a temperature difference signal
between the reactor coolant average temperature and a reference
temperature initiates rapid opening of the turbine bypass valves to
their fully-open positions. When the reactor output and the heat
removal rate from the core by the main steam system are
equilibrated, the reactor coolant temperature stops rising.
Subsequently, the temperature is decreased by automatic movements of
control rods based on the temperature difference signal, until
eventually the reactor coolant temperature is stabilized. The
pressurizer pressure increases due to an in-surge of water caused by
the reactor coolant temperature rise. Hie
pressure rise is suppressed by the actuation of the pressurizer
spray and relief valves. The coolant temperature is lowered by the
operation of turbine bypass valves, resulting in out-surging of the
pressurizer water and a decrease in the pressurizer pressure.
Finally, the system pressure is gradually restored by the
pressurizer heaters. In a standard design plant, the turbine bypass
control system can dump 40% of the rated steam flow to the condenser
and can cope with up to a 50% rapid load reduction without a reactor
trip. A house load
independent operation plant, has a design which makes if
capable of dumping 70% of the rated steam flow. This
type plant is not tripped, even if its entire off-site loads are
rejected as a result of trouble in the transmission system, and it
is switched over to the independent operation mode with only the
station house load of approximately 5%.
The main parameters to be monitored during the plant normal
operation are the reactor output (not exceeding 100%), positions of
the control rod clusters (below the upper limits of withdrawal to
maintain the controllability of the reactor and above the minimum
insertion positions determined as a function of output to keep the
shutdown margin), and the axial offset i.e. neutron flux deviation
in the axial direction (kept within the target range of +
5%).
The reactor operation with the axial offset kept within fixed limits
is called constant axial
offset control (CAOC) operation. It is adopted to keep
the reactor output distribution within limits imposed by loss of
coolant accident (LOCA) requirements. Under the CAOC operation, the
axial power distribution in the core is kept as uniform as possible
during slow load changes as seen in the case of daily
load follow operation, by minimizing the fluctuation of
axial power distribution due to xenon.
The axial offset (A*O)
is a distortion of reactor core axial power distribution,
and defined by the following equation.
Pt
—
Pb
A •
O
x 100
Pt
+
Pr
Here PT is the power of the top half of the core and PB is the power
of the bottom half. In practice AO is obtained from the upper and
lower detector signals of the four out-of-core power range channels
installed in the vicinity of the outer periphery of the reactor
vessel. Figure 5.1.4
illustrates diagrams and the locations of channels related to the
CAOC.
Though NPPs are now operated with base-loads, increased nuclear
power generation capacity in the future will require that plants are
designed to accommodate themselves to varying loads (daily load
follow operation, automatic frequency control operation, etc.) and
to continue operation in cases of transmission system accidents
(operation with only
house loads, etc.) for satisfying transmission grid needs.
Furthermore, a long-term (long-cycle) continuous operation design
for about one year or longer is now being studied to improve plant
availability factors.
NSRA,
Japan
5-8