ML17261A234
ML17261A234 | |
Person / Time | |
---|---|
Site: | Calvert Cliffs ![]() |
Issue date: | 07/19/2017 |
From: | Exelon Generation Co |
To: | Office of Nuclear Reactor Regulation |
Shared Package | |
ML17261A190 | List: |
References | |
Download: ML17261A234 (76) | |
Text
MSSVs B 3.7.1 B 3.7 PLANT SYSTEMS
B 3.7.1 Main Steam Safety Valves (MSSVs)
BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.1-1 Revision 2 BACKGROUND The primary purpose of the MSSVs is to provide overpressure protection for the secondary system. The MSSVs also provide
protection against overpressurizing the reactor coolant
pressure boundary by providing a heat sink for the removal of energy from the Reactor Coolant System (RCS) if the
preferred heat sink, provided by the condenser and Circulating Water System, is not available.
Eight MSSVs are located on each main steam header, outside the Containment Structure, upstream of the main steam isolation valves (MSIVs), as described in Reference 1, Chapter 10. The MSSV rated capacity passes the full steam flow at 102% RATED THERMAL POWER (100% + 2% for instrument error) with the valves full open. This meets the
requirements of Reference 2,Section III, Article NC-7000, Class 2 Components. The MSSV design includes staggered setpoints, according to Table 3.7.1-1 in the accompanying
Limiting Condition for Operation (LCO), so that only the number of valves needed will actuate. Staggered setpoints
reduce the potential for valve chattering, because of insufficient steam pressure to fully open all valves, following a turbine reactor trip. The MSSVs have "R" size orifices.
APPLICABLE The design basis for the MSSVs comes from Reference 2, SAFETY ANALYSES Section III, Article NC-7000, Class 2 Components; their purpose is to limit secondary system pressure to 110% of design pressure when passing 100% of design steam flow.
This design basis is sufficient to cope with any anticipated operational occurrence or accident considered Reference 1, Chapter 14.
The events that challenge the MSSV relieving capacity, and thus RCS pressure, are those characterized as decreased heat
removal events, and are presented in Reference 1, Section 14.5. Of these, the full power loss of load event is the limiting anticipated operational occurrence. A loss of load isolates the turbine and condenser, and terminates
normal feedwater flow to the steam generators. Before
delivery of auxiliary feedwater (AFW) to the steam MSSVs B 3.7.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.1-2 Revision 61 generators, RCS pressure reaches peak pressure. The peak pressure is < 110% of the design pressure of 2500 psia, but
high enough to actuate the pressurizer safety valves.
Although the Power Level-High Trip is not credited in the loss of load safety analysis, reducing the Power Level-High Trip setpoint ensures the Thermal Power limit supported by the safety analysis is met.
The MSSVs satisfy 10 CFR 50.36(c)(2)(ii), Criterion 3.
LCO This LCO requires all MSSVs to be OPERABLE in compliance with Reference 2,Section III, Article NC-7000, Class 2
Components, even though this is not a requirement of the
Design Basis Accident (DBA) analysis. This is because
operation with less than the full number of MSSVs requires
limitations on allowable THERMAL POWER (to meet Reference 2,Section III, Article NC-7000, Class 2 Components
requirements), and adjustment to the Reactor Protective
System trip setpoints to meet the transient analysis limits.
These limitations are according to those shown in
Table 3.7.1-1, Required Action A.2, and Required Action A.3
in the accompanying LCO.
The OPERABILITY of the MSSVs is defined as the ability to open within the setpoint tolerances, relieve steam generator
overpressure, and reseat when pressure has been reduced.
The OPERABILITY of the MSSVs is determined by periodic
surveillance testing in accordance with the INSERVICE TESTING PROGRAM. An MSSV is considered inoperable if it fails to open upon demand.
The lift settings, according to Table 3.7.1-2 in the accompanying LCO, correspond to ambient conditions of the
valve at nominal operating temperature and pressure.
A Note is added to Table 3.7.1-2, stating that lift settings for a given steam line are also acceptable, if any two
valves lift between 935 and 1005 psig, any two other valves
lift between 935 and 1035 psig, and the four remaining
valves lift between 935 and 1050 psig. Thus, the MSSVs
still perform that design basis function properly.
MSSVs B 3.7.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.1-3 Revision 23 This LCO provides assurance that the MSSVs will perform their designed safety function to mitigate the consequences
of accidents that could result in a challenge to the reactor
coolant pressure boundary.
APPLICABILITY In MODEs 1, 2, and 3, a minimum of five MSSVs per steam generator are required to be OPERABLE, according to
Table 3.7.1-1 in the accompanying LCO, which is limiting and
bounds all lower MODEs.
In MODEs 4 and 5, there are no credible transients requiring the MSSVs.
The steam generators are not normally used for heat removal in MODEs 5 and 6, and thus cannot be overpressurized; there
is no requirement for the MSSVs to be OPERABLE in these MODEs. ACTIONS The ACTIONS table is modified by a Note indicating that separate Condition entry is allowed for each MSSV.
A.1 and A.2 An alternative to restoring the inoperable MSSV(s) to OPERABLE status is to reduce power so that the available
MSSV relieving capacity meets Code requirements for the power level. The number of inoperable MSSVs will determine the necessary level of reduction in secondary system steam
flow and THERMAL POWER required by the reduced reactor trip
settings of the power level-high channels. The setpoints in
Table 3.7.1-1 have been verified by transient analyses.
The operator should limit the maximum steady state power level to some value slightly below this setpoint to avoid an
inadvertent overpower trip.
The four-hour Completion Time for Required Action A.1 is a reasonable time period to reduce power level and is based on
the low probability of an event occurring during this period
that would require activation of the MSSVs. An additional
32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> is allowed in Required Action A.2 to reduce the setpoints. The Completion Time of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> for Required Action A.2 is based on a reasonable time to correct the MSSV MSSVs B 3.7.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.1-4 Revision 61 inoperability, the time required to perform the power reduction, operating experience in resetting all channels of
a protective function, and on the low probability of the
occurrence of a transient that could result in steam
generator overpressure during this period.
B.1 and B.2 If the MSSVs cannot be restored to OPERABLE status in the associated Completion Time, or if one or more steam
generators have less than five MSSVs OPERABLE, the unit must
be placed in a MODE in which the LCO does not apply. To
achieve this status, the unit must be placed in at least
MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The
allowed Completion Times are reasonable, based on operating
experience, to reach the required unit conditions from full
power conditions in an orderly manner and without challenging unit systems.
SURVEILLANCE SR 3.7.1.1 REQUIREMENTS This Surveillance Requirement (SR) verifies the OPERABILITY of the MSSVs by the verification of each MSSV lift setpoints
in accordance with the INSERVICE TESTING PROGRAM. The safety and relief valve tests are to be performed in
accordance with Reference 3. According to Reference 3, the
following tests are required for MSSVs: a. Visual examination;
- b. Seat tightness determination;
- c. Setpoint pressure determination (lift setting);
- d. Compliance with owner's seat tightness criteria; and e. Verification of the balancing device integrity on balanced valves.
The ANSI/American Society of Mechanical Engineers (ASME)
Standard requires that all valves be tested every
five years, and a minimum of 20% of the valves be tested
every 24 months. The ASME Code specifies the activities, as
found lift acceptance range, and frequencies necessary to
satisfy the requirements. Table 3.7.1-2 defines the lift
setting range for each MSSV for OPERABILITY; however, the MSSVs B 3.7.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.1-5 Revision 38 valves are reset to + 1% during the surveillance test to allow for drift.
This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. This is to
allow testing of the MSSVs at hot conditions. The MSSVs may be either bench tested or tested in situ at hot conditions, using an assist device to simulate lift pressure. If the
MSSVs are not tested at hot conditions, the lift setting
pressure shall be corrected to ambient conditions of the valve at operating temperature and pressure.
REFERENCES 1. Updated Final Safety Analysis Report (UFSAR) 2. ASME, Boiler and Pressure Vessel Code
- 3. ANSI/ASME OM-1-1987, Code for the Operation and Maintenance of Nuclear Power Plants, 1987
MSIVs B 3.7.2 B 3.7 PLANT SYSTEMS
B 3.7.2 Main Steam Isolation Valves (MSIVs)
BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.2-1 Revision 14 BACKGROUND The MSIVs isolate steam flow from the secondary side of the steam generators following a high energy line break (HELB).
Main steam isolation valve closure terminates flow from the
unaffected (intact) steam generator.
One MSIV is located in each main steam line outside, but close to, the Containment Structure. The MSIVs are
downstream from the MSSVs, atmospheric dump valves (ADVs),
and AFW pump turbine steam supplies to prevent their being
isolated from the steam generators by MSIV closure. Closing
the MSIVs isolates each steam generator from the other, and
isolates the turbine, Steam Bypass System, and other
auxiliary steam supplies from the steam generators.
The MSIVs close on a steam generator isolation signal generated by low steam generator pressure or on a
containment spray actuation signal (CSAS) generated by high
containment pressure. The MSIVs fail closed on loss of
control or actuation power. The steam generator isolation
signal also actuates the main feedwater isolation valves (MFIVs) to close. The MSIVs may also be actuated manually.
A description of the MSIVs is found in Reference 1, Section 10.1.
APPLICABLE The design basis of the MSIVs is established by the SAFETY ANALYSES containment analysis for the large steam line break (SLB) inside the Containment Structure, as discussed in Reference 1, Section 14.20. It is also influenced by the accident analysis of the SLB events presented in
Reference 1, Section 14.14. The design precludes the
blowdown of more than one steam generator, assuming a single
active component failure (e.g., the failure of one MSIV to
close on demand).
The limiting case for main SLB Containment Structure response is 75% power, no loss of offsite power, and failure of a steam generator feed pump to trip. This case results in continued feeding of the affected steam generator and
maximizes the energy release into the Containment Structure.
MSIVs B 3.7.2 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.2-2 Revision 14 This case does not assume failure of an MSIV; however, an important assumption is both MSIVs are OPERABLE. This
prevents blowdown of both steam generators assuming failure
of an MSIV to close.
The accident analysis compares several different SLB events against different acceptance criteria. The large SLB outside the Containment Structure upstream of the MSIV is
the limiting SLB for offsite dose, although a break in this
short section of main steam header has a very low
probability. The large SLB inside the Containment Structure
at hot full power is the limiting case for a post-trip
return to power. The analysis includes scenarios with
offsite power available and with a loss of offsite power
following turbine trip.
The MSIVs only serve a safety function and remain open during power operation. These valves operate under the
following situations: a. An HELB inside the Containment Structure. In order to maximize the mass and energy release into the
Containment Structure, the analysis assumes steam is
discharged into the Containment Structure from both
steam generators until closure of the MSIV occurs.
After MSIV closure, steam is discharged into the
Containment Structure only from the affected steam
generator. b. A break outside of the Containment Structure and upstream from the MSIVs. This scenario is not a
containment pressurization concern. The uncontrolled
blowdown of more than one steam generator must be
prevented to limit the potential for uncontrolled RCS
cooldown and positive reactivity addition. Closure of
the MSIVs limits the blowdown to a single steam generator. c. A break downstream of the MSIVs. This type of break will be isolated by the closure of the MSIVs. Events
such as increased steam flow through the turbine or the
steam bypass valves (e.g., excess load event) will also
terminate on closure of the MSIVs. d. A steam generator tube rupture. For this scenario, closure of the MSIV isolates the affected steam MSIVs B 3.7.2 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.2-3 Revision 14 generator from the intact steam generator and minimizes radiological releases. The operator is then required
to maintain the pressure of the steam generator with
the ruptured tube below the MSSV setpoints, a necessary
step toward isolating the flow through the rupture. e. The MSIVs are also utilized during other events such as a feedwater line break. These events are less limiting so far as MSIV OPERABILITY is concerned.
The MSIVs satisfy 10 CFR 50.36(c)(2)(ii), Criterion 3.
LCO This LCO requires that the MSIV in each of the two steam lines be OPERABLE. The MSIVs are considered OPERABLE when
the isolation times are within limits, and they close on an
isolation actuation signal.
This LCO provides assurance that the MSIVs will perform their design safety function to mitigate the consequences of accidents as described in Reference 1, Chapter 14.
APPLICABILITY The MSIVs must be OPERABLE in MODE 1 and in MODEs 2 and 3, except when all MSIVs are closed. In these MODEs there is
significant mass and energy in the RCS and steam generators.
When the MSIVs are closed, they are already performing their
safety function.
In MODE 4, the steam generator energy is low; therefore, the MSIVs are not required to be OPERABLE.
In MODEs 5 and 6, the steam generators do not contain much energy because their temperature is below the boiling point
of water; therefore, the MSIVs are not required for
isolation of potential high energy secondary system pipe breaks in these MODEs.
ACTIONS A.1 With one MSIV inoperable in MODE 1, time is allowed to restore the component to OPERABLE status. Some repairs can
be made to the MSIV with the unit hot. The eight hour
Completion Time is reasonable, considering the probability
of an accident occurring during the time period that would
require closure of the MSIVs.
MSIVs B 3.7.2 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.2-4 Revision 14 B.1 If the MSIV cannot be restored to OPERABLE status within eight hours, the unit must be placed in a MODE in which the
LCO does not apply. To achieve this status, the unit must
be placed in MODE 2 within six hours and Condition C would be entered. The Completion Time is reasonable, based on operating experience, to reach MODE 2, and close the MSIVs
in an orderly manner and without challenging unit systems.
C.1 and C.2 Condition C is modified by a Note indicating that separate Condition entry is allowed for each MSIV.
Since the MSIVs are required to be OPERABLE in MODEs 2 and 3, the inoperable MSIVs may either be restored to
OPERABLE status or closed. When closed, the MSIVs are
already in the position required by the assumptions in the
safety analysis.
The eight hour Completion Time is consistent with that allowed in Condition A.
Inoperable MSIVs that cannot be restored to OPERABLE status within the specified Completion Time, but are closed, must
be verified on a periodic basis to be closed. This is
necessary to ensure that the assumptions in the safety
analysis remain valid. The seven day Completion Time is
reasonable, based on engineering judgment, MSIV status
indications available in the Control Room, and other
administrative controls, to ensure these valves are in the
closed position.
D.1 and D.2 If the MSIVs cannot be restored to OPERABLE status, or closed, within the associated Completion Time, the unit must
be placed in a MODE in which the LCO does not apply. To
achieve this status, the unit must be placed in at least
MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The
allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from MSIVs B 3.7.2 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.2-5 Revision 61 MODE 2 conditions in an orderly manner and without challenging unit systems.
SURVEILLANCE SR 3.7.2.1 REQUIREMENTS This SR verifies that the closure time of each MSIV is
< 5.2 seconds. The MSIV closure time is assumed in the accident and containment analyses.
The Frequency for this SR is in accordance with the INSERVICE TESTING PROGRAM. The MSIVs are tested during each refueling outage in accordance with Reference 2, and
sometimes during other cold shutdown periods. The Frequency
demonstrates the valve closure time at least once per
refueling cycle. Operating experience has shown that these
components usually pass the SR when performed. Therefore, the Frequency is acceptable from a reliability standpoint.
REFERENCES 1. UFSAR 2. ASME Code for Operation and Maintenance of Nuclear Power Plants
AFW System B 3.7.3 B 3.7 PLANT SYSTEMS
B 3.7.3 Auxiliary Feedwater (AFW) System
BASES CALVERT CLIFFS - UNIT 1 & 2 B 3.7.3-1 Revision 2 BACKGROUND The AFW System automatically supplies feedwater to the steam generators to remove decay heat from the RCS upon the loss of normal feedwater supply. The AFW pumps take suction
through a common suction line from the condensate storage tank (CST) (LCO 3.7.4) and pump to the steam generator secondary side via separate and independent connections, to the AFW header outside the Containment Structure. The steam generators function as a heat sink for core decay heat. The heat load is dissipated by releasing steam to the atmosphere
from the steam generators via the MSSVs (LCO 3.7.1) or ADVs.
If the main condenser is available, steam may be released
via the steam bypass valves and the resulting excess water
inventory in the hotwell is moved to the backup water
supply.
The AFW System consists of, one motor-driven AFW pump and two steam turbine-driven pumps configured into two trains.
The motor-driven pump provides 100% of AFW flow capacity; each turbine-driven pump can provide 100% of the required capacity to the steam generators as assumed in the accident
analysis, but only one turbine-driven pump is lined up to auto start. The other turbine-driven pump is placed in standby and requires a manual start, when it is needed. The pumps are equipped with a common recirculation line to
prevent pump operation against a closed system. The motor-driven AFW pump is powered from an independent Class 1E
power supply, and feeds both steam generators.
One pump at full flow is sufficient to remove decay heat and cool the unit to Shutdown Cooling (SDC) System entry
conditions.
The steam turbine-driven AFW pumps receive steam from either main steam header upstream of the MSIV. Each of the steam feed lines will supply 100% of the requirements of the turbine-driven AFW pump. The turbine-driven AFW pump supplies a common header capable of feeding both steam
generators, with air-operated valves (with controllers powered by AC vital buses) actuated to the appropriate steam AFW System B 3.7.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.3-2 Revision 60 generator by the Auxiliary Feedwater Actuation System (AFAS).
The AFW System may also supply feedwater to the steam generators during normal unit startup, shutdown, and hot
standby conditions although the normal supply is main feedwater (MFW).
The AFW System is designed to supply sufficient water to the steam generator(s) to remove decay heat with steam generator
pressure at the setpoint of the MSSVs. Subsequently, the
AFW System supplies sufficient water to cool the unit to SDC
entry conditions, and steam is released through the ADVs.
The AFW System actuates automatically on low steam generator level by the AFAS, as described in LCO 3.3.4. The AFAS
logic is designed to feed either or both steam generators
with low levels, but will isolate the AFW System from a
steam generator having a significantly lower steam pressure
than the other steam generator. The AFAS automatically
actuates one AFW turbine-driven pump and associated air-
operated valves (with controllers powered by AC vital buses)
when required, to ensure an adequate feedwater supply to the
steam generators. Air-operated valves with controllers
powered by AC vital busses are provided for each AFW line to
control the AFW flow to each steam generator.
The AFW System is discussed in Reference 1.
APPLICABLE The AFW System mitigates the consequences of any event with SAFETY ANALYSES a loss of normal feedwater.
The design basis of the AFW System is to supply water to the steam generator to remove decay heat and other residual
heat, by delivering at least the minimum required flow rate
to the steam generators at pressures corresponding to the
lowest MSSV set pressure plus 3%.
The limiting DBAs and transients for the AFW System are as follows: a. Main SLB; b. Loss of normal feedwater; and AFW System B 3.7.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.3-3 Revision 60 c. Feedwater Line Break.
The AFW System satisfies 10 CFR 50.36(c)(2)(ii), Criterion 3.
LCO This LCO requires that two AFW trains be OPERABLE to ensure that the AFW System will perform its design safety function.
A train consists of one pump and the piping, valves, and
controls in the direct flow path. Three AFW pumps are
installed, consisting of one motor-driven and two non-
condensing steam turbine-driven pumps. For a shutdown, only
one pump is required to be operating, the others are in
standby. Upon automatic initiation of AFW, one motor-driven
and one turbine-driven pump automatically start.
The AFW System is considered to be OPERABLE when the components and flow paths required to provide AFW flow to
the steam generators are OPERABLE. This requires that the
motor-driven AFW pump be OPERABLE and capable of supplying
AFW flow to both steam generators. The turbine-driven AFW
pumps shall be OPERABLE with redundant steam supplies from
each of the two main steam lines upstream of the MSIVs and
capable of supplying AFW flow to both of the two steam
generators. The piping, valves, instrumentation, and
controls in the required flow paths shall also be OPERABLE.
The LCO is modified by a Note that allows AFW trains required for Operability to be taken out-of-service under
administrative control for the performance of periodic
testing. This LCO note allows a limited exception to the LCO requirement and allows this condition to exist without requiring any Technical Specification Condition to be
entered. The following administrative controls are
necessary during periodic testing to ensure the operator(s)
can restore the AFW train(s) from the test configuration to
its operational configuration when required. A dedicated
operator(s) is stationed at the control station(s) with
direct communication to the Control Room whenever the
train(s) is in the testing configuration. Upon completion
of the testing the trains are returned to proper status and
verified in proper status by independent operator checks.
The administrative controls include certain operator
restoration actions that are virtually certain to be AFW System B 3.7.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.3-4 Revision 60 successful during accident conditions. These actions include but are not limited to the following: operation of
pump discharge valves, operation of trip/throttle valve(s),
simple handswitch/controller manipulations, and adjusting
the local governor speed control knob. The administrative
controls do not include actions to restore a tripped AFW pump due to the complicated nature of this task. Periodic tests include those tests that are performed in a controlled
manner similar to surveillance tests, but not necessarily on
the established surveillance test schedule, such as post-
maintenance tests. This Note is necessary because of the
AFW pump configuration.
APPLICABILITY In MODEs 1, 2, and 3, the AFW System is required to be OPERABLE and to function in the event that the MFW is lost.
In addition, the AFW System is required to supply enough
makeup water to replace steam generator secondary inventory
and maintain the RCS in MODE 3.
In MODE 4, the AFW System is not required, however, it may be used for heat removal via the steam generator although
the preferred method is MFW.
In MODEs 5 and 6, the steam generators are not normally used for decay heat removal, and the AFW System is not required.
ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable AFW train. There is an increased risk associated
with entering a MODE or other specified condition in the
Applicability with an AFW train inoperable and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO
not met after performance of a risk assessment addressing
inoperable systems and components, should not be applied in
this circumstance.
A.1 and A.2 With one of the required steam-driven AFW pumps inoperable, action must be taken to align the remaining OPERABLE steam-
driven pump to automatic initiating status. This Required
Action ensures that a steam-driven AFW pump is available to
automatically start, if required. If the OPERABLE AFW pump
is properly aligned, the inoperable steam-driven AFW pump AFW System B 3.7.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.3-5 Revision 60 must be restored to OPERABLE status (and placed in either standby or automatic initiating status, depending upon
whether the other steam-driven AFW pump is in standby or
automatic initiating status) within seven days. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
and seven day Completion Times are reasonable, based on the
redundant capabilities afforded by the AFW System, the time needed for repairs, and the low probability of a DBA event occurring during this period. Two AFW pumps and flow paths
remain to supply feedwater to the steam generators.
B.1 and B.2 With the motor-driven AFW pump inoperable, action must be taken to align the standby steam-driven pump to automatic
initiating status. This Required Action ensures that
another AFW pump is available to automatically start, if
required. If the standby steam-driven pump is properly
aligned, the inoperable motor-driven AFW pump must be
restored to OPERABLE status within seven days. The 72-hour
and seven day, Completion Times are reasonable, based on the
redundant capabilities afforded by the AFW System, the time
needed for repairs, and the low probability of a DBA event
occurring during this period. Two AFW pumps and one flow
path remain to supply feedwater to the steam generators.
C.1, C.2, C.3, and C.4 With two AFW pumps inoperable, action must be taken to align the remaining OPERABLE pump to automatic initiating status
and to verify the other units motor-driven AFW pump is
OPERABLE, along with an OPERABLE cross-tie valve, within
one hour. If these Required Actions are completed within
the Completion Time, one AFW pump must be restored to
OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Verifying the other unit's
motor-driven AFW pump is OPERABLE provides an additional
level of assurance that AFW will be available if needed, because the other unit's AFW can be cross-connected if necessary. The cross-tie valve to the opposite unit is
administratively verified OPERABLE by confirming that
SR 3.7.3.2 has been performed within the specified
Frequency. These one hour Completion Times are reasonable
based on the low probability of a DBA occurring during the
first hour and the need for AFW during the first hour. The
72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> completion time to restore one AFW pump to OPERABLE AFW System B 3.7.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.3-6 Revision 60 status takes into account the cross-connected capability between units and the unlikelihood of an event occurring in
the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period.
D.1 With one of the required AFW trains inoperable for reasons other than Condition A, B, or C (e.g., flowpath or steam supply valve), action must be taken to restore OPERABLE
status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. This Condition includes the loss of
two steam supply lines to the turbine-driven AFW pumps. The
72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is reasonable, based on the
redundant capabilities afforded by the AFW System, the time
needed for repairs, and the low probability of a DBA event
occurring during this period. One AFW train remains to
supply feedwater to the steam generators.
E.1 and E.2 When the Required Action and associated Completion Time of Condition A, B, C, or D cannot be met the unit must be
placed in a MODE in which the LCO does not apply. To
achieve this status, the unit must be placed in at least
MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions
from full power conditions in an orderly manner and without
challenging unit systems.
F.1 Required Action F.1 is modified by a Note indicating that all required MODE changes or power reductions are suspended
until one AFW train is restored to OPERABLE status.
With two AFW trains inoperable in MODEs 1, 2, and 3, the unit may be in a seriously degraded condition with only non-safety-related means for conducting a cooldown. In such a
condition, the unit should not be perturbed by any action, including a power change, that might result in a trip.
However, a power change is not precluded if it is determined
to be the most prudent action. The seriousness of this
condition requires that action be started immediately to
restore one AFW train to OPERABLE status. While other plant AFW System B 3.7.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.3-7 Revision 61 conditions may require entry into LCO 3.0.3, the ACTIONS required by LCO 3.0.3 do not have to be completed because
they could force the unit into a less safe condition.
SURVEILLANCE SR 3.7.3.1 REQUIREMENTS Verifying the correct alignment for manual, power-operated, and automatic valves in the AFW water and steam supply flow
paths, provides assurance that the proper flow paths exist
for AFW operation. This SR does not apply to valves that
are locked, sealed, or otherwise secured in position, since
these valves are verified to be in the correct position
prior to locking, sealing, or securing. This SR also does
not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing
or valve manipulations; rather, it involves verification
that those valves capable of potentially being mispositioned
are in the correct position.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.7.3.2 Cycling each testable, remote-operated valve that is not in its operating position, provides assurance that the valves
will perform as required. Operating position is the
position that the valve is in during normal plant operation.
This is accomplished by cycling each valve at least one
cycle. This SR ensures that valves required to function
during certain scenarios, will be capable of being properly
positioned. The Frequency is based on engineering judgment
that when cycled in accordance with the INSERVICE TESTING PROGRAM, these valves can be placed in the desired position when required.
SR 3.7.3.3 Verifying that each AFW pump's developed head at the flow test point is greater than or equal to the required developed head ( 2800 ft for the steam-driven pump and 3100 ft for the motor-driven pump), ensures that AFW pump performance has not degraded during the cycle. Flow and
differential head are normal tests of pump performance AFW System B 3.7.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.3-8 Revision 60 required by Reference 2. Because it is undesirable to introduce cold AFW into the steam generators while they are
operating, this testing is performed on recirculation flow.
This test confirms one point on the pump design curve and is
indicative of overall performance. Such inservice tests
confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance.
Performance of inservice testing, discussed in Reference 2, at three month intervals satisfies this requirement.
This SR is modified by a Note indicating that the SR should be deferred up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until suitable test conditions
are established. This deferral is required because there is
an insufficient steam pressure to perform the test.
SR 3.7.3.4 This SR ensures that AFW can be delivered to the appropriate steam generator, in the event of any accident or transient
that generates an AFAS signal, by demonstrating that each
automatic valve in the flow path actuates to its correct
position on an actual or simulated actuation signal (verification of flow-modulating characteristics is not
required). This SR is not required for valves that are
locked, sealed, or otherwise secured in the required
position under administrative controls. The Surveillance
Frequency is controlled under the Surveillance Frequency
Control Program.
This SR is modified by a Note indicating that the SR should be deferred up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until suitable test conditions
have been established.
SR 3.7.3.5 This SR ensures that the AFW pumps will start in the event of any accident or transient that generates an AFAS signal by demonstrating that each AFW pump starts automatically on
an actual or simulated actuation signal. The Surveillance
Frequency is controlled under the Surveillance Frequency
Control Program.
AFW System B 3.7.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.3-9 Revision 60 This SR is modified by a Note. The Note indicates that the SR should be deferred up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until suitable test
conditions are established.
SR 3.7.3.6 This SR ensures that the AFW system is capable of providing a minimum nominal flow to each flow leg. This ensures that the minimum required flow is capable of feeding each flow
leg. The test may be performed on one flow leg at a time.
The SR is modified by a Note which states, the SR is not
required to be performed for the AFW train with the turbine-
driven AFW pump until up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 800 psig
in the steam generators. The Note ensures that proper test
conditions exist prior to performing the test using the
turbine-driven AFW pumps. The Surveillance Frequency is
controlled under the Surveillance Frequency Control Program.
SR 3.7.3.7 This SR ensures that the AFW System is properly aligned by verifying the flow path to each steam generator prior to
entering MODE 2 operation, after 30 days in MODEs 5 or 6.
OPERABILITY of AFW flow paths must be verified before
sufficient core heat is generated that would require the
operation of the AFW System during a subsequent shutdown.
The Frequency is reasonable, based on engineering judgment, and other administrative controls to ensure that flow paths
remain OPERABLE. To further ensure AFW System alignment, the OPERABILITY of the flow paths is verified following
extended outages to determine that no misalignment of valves
has occurred. This SR ensures that the flow path from the
CST to the steam generators is properly aligned. Minimum
nominal flow to each flow leg is ensured by performance of SR 3.7.3.6.
REFERENCES 1. UFSAR, Section 10.3 2. ASME Code for Operation and Maintenance of Nuclear Power Plants
CST B 3.7.4 B 3.7 PLANT SYSTEMS
B 3.7.4 Condensate Storage Tank (CST)
BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.4-1 Revision 41 BACKGROUND The CST provides a safety grade source of water to the steam generators for removing decay and sensible heat from the
RCS. The CST provides a passive flow of water, by gravity, to the AFW System (LCO 3.7.3). The steam produced is released to the atmosphere by the MSSVs or the atmospheric dump valves. The AFW pumps operate with a continuous
recirculation to the CST.
The component required by this Specification is CST No. 12.
When the MSIVs are open, the preferred means of heat removal is to discharge steam to the condenser by the non-safety
grade path of the turbine bypass valves. The condensed
steam is returned to the backup water supply (CST No. 11 and
CST No. 21) by the condensate pump. This has the advantage
of conserving condensate while minimizing releases to the
environment.
Because the CST is a principal component in removing residual heat from the RCS, it is designed to withstand
earthquakes and other natural phenomena. The CST is
designed to Seismic Category I requirements to ensure
availability of the feedwater supply. Feedwater is also
available from an alternate source.
There is one CST (CST No. 12) shared by Units 1 and 2. A description of the CST is found in Reference 1, Sections 6.3.5.1 and 10.3.2.
APPLICABLE The CST provides cooling water to remove decay heat and to SAFETY ANALYSES cool down the unit following all events except for the maximum hypothetical accident and the fuel handling accident in the accident analyses, discussed in Reference 1, Chapter 14. For anticipated operational occurrences and
accidents which do not affect the OPERABILITY of the steam
generators, the thermal analysis assumption is generally six hours at MODE 3, steaming through the ADVs and MSSVs followed by a cooldown to SDC entry conditions at the design
cooldown rate. The dose analysis assumption is an eight hour cooldown to maximize Control Room and offsite doses.
CST B 3.7.4 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.4-2 Revision 41 The limiting event for the condensate volume is the large feedwater line break with a coincident loss of offsite
power. Single failures that also affect this event include
the following: a. The failure of the diesel generator powering the motor-driven AFW pump to the unaffected steam generator (requiring additional steam to drive the remaining AFW
pump turbine); and b. The failure of the steam driven train (requiring a longer time for cooldown using only one motor-driven AFW pump).
These are not usually the limiting failures in terms of consequences for these events.
The CST satisfies 10 CFR 50.36(c)(2)(ii), Criteria 2 and 3.
LCO To satisfy accident analysis assumptions, CST No. 12 must contain sufficient cooling water for both units to ensure
that sufficient water is available to maintain the RCS at
MODE 3 for six hours following a reactor trip from
102% RATED THERMAL POWER, assuming a coincident loss of
offsite power and the most adverse single failure. In doing
this, it must retain sufficient water to ensure adequate net
positive suction head for the AFW pumps during the cooldown
while in MODE 3, as well as to account for any losses from
the steam-driven AFW pump turbine, or before isolating AFW
to a broken line.
The CST usable volume required is 150,000 gallons per unit (300,000 gallons for both units) in the MODE of
Applicability. The 300,000 gallons of water is enough to
provide for decay heat removal and cooldown of both units.
By adjusting the feedwater flow to the permissible cooldown rate, decay heat removal and cooldown of both units can be accomplished in six hours. The 300,000 gallons are also
adequate to maintain the RCS in MODE 3 for six hours with
steam discharge to atmosphere with concurrent and total loss
of offsite power, or to remove decay heat from both units
for more than ten hours after initiation of cooldown and
still maintain normal no-load water level in the steam
generators. The total water volume in the tank includes the CST B 3.7.4 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.4-3 Revision 41 usable volume and water not usable because of the tank discharge line location.
OPERABILITY of the CST is determined by maintaining the tank volume at or above the minimum required volume.
APPLICABILITY In MODEs 1, 2, and 3, the CST is required to be OPERABLE.
In MODEs 4, 5 and 6, the CST is not required because the AFW System is not required.
ACTIONS A.1 and A.2 If the CST is not OPERABLE, the OPERABILITY of the backup water supply (CST No. 11 for Unit 1 and CST No. 21 for
Unit 2) must be verified by administrative means within
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.
OPERABILITY of the backup feedwater supply must include verification that the manual valves in the flow paths from
the backup supply to the AFW pumps are open, and
availability of the required volume of water
(150,000 gallons) in the backup supply. The CST must be
returned to OPERABLE status within seven days, as the backup
supply may be performing this function in addition to its
normal functions. The four hour Completion Time is
reasonable, based on operating experience, to verify the
OPERABILITY of the backup water supply. Additionally, verifying the backup water supply every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is adequate
to ensure the backup water supply continues to be available.
The seven day Completion Time is reasonable, based on an OPERABLE backup water supply being available, and the low probability of an event requiring the use of the water from
the CST occurring during this period.
If the CST volume is less than 300,000 gallons and greater than 150,000 gallons and both units are in the MODE of
Applicability, only one unit must enter this condition
provided the unit aligns to the OPERABLE backup water supply (CST No. 11 or CST No. 21).
CST B 3.7.4 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.4-4 Revision 55 B.1 and B.2 If the CST cannot be restored to OPERABLE status within the associated Completion Time, the affected unit(s) must be
placed in a MODE in which the LCO does not apply. To
achieve this status, the unit(s) must be placed in at least
MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The
allowed Completion Times are reasonable, based on operating
experience, to reach the required unit conditions from full
power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.7.4.1 REQUIREMENTS This SR verifies that the CST contains the required usable volume of cooling water. (This volume 150,000 gallons per unit in the MODE of Applicability.) The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Although the volume in the CST for each unit is required to be 150,000 gallons, the total combined volume for both units is 300,000 gallons.
REFERENCES 1. UFSAR
CC System B 3.7.5 B 3.7 PLANT SYSTEMS
B 3.7.5 Component Cooling (CC) System
BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.5-1 Revision 53 BACKGROUND The CC System provides a heat sink for the removal of process and operating heat from safety-related components
during a DBA or transient. During normal operation, the CC
System also provides this function for various nonessential components. The CC System serves as a barrier to the release of radioactive byproducts between potentially
radioactive systems and the Saltwater (SW) System, and thus
to the environment.
The CC System consists of two redundant loops that are always cross-connected. A loop consists of one of three
redundant pumps, one of two redundant CC heat exchangers
along with a common head tank, associated valves, piping, instrumentation, and controls. The third pump, which is an
installed spare, can be powered from either electrical
train. The redundant cooling capacity of this system, assuming single active failure, is consistent with the
assumptions made in the accident analysis.
During normal operation one loop typically provides cooling water with a maximum CC heat exchanger outlet temperature of 95°F (a range of 70
°F-95°F is acceptable during normal operating conditions) with the redundant loop components in
standby. If needed, the redundant loop components can be
aligned to supplement the in service loop. While operating
on SDC with one loop, the CC heat exchanger outlet temperature may rise to a maximum temperature of 120
°F. Additional information on the design and operation of the system, along with a list of the components served, is
presented in Reference 1, Section 9.5.2.1. The principal
safety-related function of the CC System is the removal of
decay heat from the reactor via the SDC System heat exchanger. This may utilize the SDC heat exchanger, during a normal or post accident cooldown and shutdown, or the
Containment Spray System during the recirculation phase following a loss of coolant accident (LOCA).
CC System B 3.7.5 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.5-2 Revision 53 APPLICABLE The design basis of the CC System is for it to support a SAFETY ANALYSES 100% capacity Containment Cooling System (containment spray, containment coolers, or a combination) removing core decay
heat 30 minutes after a design basis LOCA. This prevents
the containment sump fluid from increasing in temperature
during the recirculation phase following a LOCA, and provides a gradual reduction in the temperature of this fluid as it is supplied to the RCS by the safety injection
pumps.
The CC System is designed to perform its function with a single failure of any active component, assuming a loss of
offsite power.
The CC System also functions to cool the unit from SDC entry conditions (Tcold < 300°F) to Tcold < 140°F during normal operations. The time required to cool from 300°F to 140°F
is a function of the number of CC and SDC loops operating.
One CC loop is sufficient to remove decay heat during
subsequent operations with Tcold < 140°F. This assumes that a maximum inlet SW temperature occurs simultaneously with
the maximum heat loads on the system.
The CC System satisfies 10 CFR 50.36(c)(2)(ii), Criterion 3.
LCO The CC loops are redundant of each other to the degree that each has separate controls and power supplies and the
operation of one does not depend on the other. In the event
of a DBA, one CC loop is required to provide the minimum
heat removal capability assumed in the safety analysis for the systems to which it supplies cooling water. To ensure this requirement is met, two CC loops must be OPERABLE. At
least one CC loop will operate assuming the worst single
active failure occurs coincident with the loss of offsite
power. Additionally, the containment cooling function will
also operate assuming the worst case passive failure post-recirculation actuation signal (RAS).
A CC loop is considered OPERABLE when the following: a. The associated pump and common head tank are OPERABLE; and CC System B 3.7.5 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.5-3 Revision 53 b. The associated piping, valves, heat exchanger and instrumentation and controls required to perform the
safety-related function are OPERABLE.
The isolation of CC from other components or systems not required for safety may render those components or systems inoperable, but does not affect the OPERABILITY of the CC
System. APPLICABILITY In MODEs 1, 2, 3, and 4, the CC System is a normally operating system that must be prepared to perform its post
accident safety functions, primarily RCS heat removal by
cooling the SDC heat exchanger.
In MODEs 5 and 6, the OPERABILITY requirements of the CC System are determined by the systems it supports.
ACTIONS A.1 Required Action A.1 is modified by a Note indicating the requirement of entry into the applicable Conditions and
Required Actions of LCO 3.4.6, for SDC made inoperable by
CC. This is an exception to LCO 3.0.6 and ensures the
proper actions are taken for these components.
With one CC loop inoperable, action must be taken to restore OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In this Condition, the remaining OPERABLE CC loop is adequate to perform the heat
removal function. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is based on
the redundant capabilities afforded by the OPERABLE loop, and the low probability of a DBA occurring during this
period.
B.1 and B.2 If the CC loop cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a
MODE in which the LCO does not apply. To achieve this
status, the unit must be placed in at least MODE 3 within
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions CC System B 3.7.5 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.5-4 Revision 55 from full power conditions in an orderly manner and without challenging unit systems.
SURVEILLANCE SR 3.7.5.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the CC flow path provides assurance
that the proper flow paths exist for CC operation. This SR
does not apply to valves that are locked, sealed, or
otherwise secured in position, since these valves are
verified to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to valves
that cannot be inadvertently misaligned, such as check
valves. This SR does not require any testing or valve
manipulation; rather, it involves verification that those
valves capable of potentially being mispositioned are in
their correct position.
This SR is modified by a Note indicating that the isolation of the CC components or systems may render those components
inoperable but does not affect the OPERABILITY of the CC
System.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.7.5.2 This SR verifies proper automatic operation of the CC valves on an actual or simulated safety injection actuation signal (SIAS). The CC System is a normally operating system that cannot be fully actuated as part of routine testing during normal operation. This SR is not required for valves that
are locked, sealed, or otherwise secured in the required
position under administrative controls. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.7.5.3 This SR verifies proper automatic operation of the CC pumps on an actual or simulated SIAS. The CC System is a normally
operating system that cannot be fully actuated as part of
routine testing during normal operation. The Surveillance CC System B 3.7.5 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.5-5 Revision 55 Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES 1. UFSAR
SRW System B 3.7.6 B 3.7 PLANT SYSTEMS
B 3.7.6 Service Water (SRW) System
BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.6-1 Revision 5 BACKGROUND The SRW System provides a heat sink for the removal of process and operating heat from safety-related components
during a DBA or transient. During normal operation or a
normal shutdown, the SRW System also provides this function for various safety-related and non-safety-related components. The safety-related function is covered by this
LCO.
The SRW System consists of two separate, 100% capacity safety-related cooling water subsystems. Each subsystem
consists of a 100% capacity pump, head tank, two SRW heat exchangers, piping, valves, and instrumentation. A third pump, which is an installed spare, can be powered from
either electrical train. The pumps and valves are remote
manually aligned, except in the unlikely event of a LOCA.
The pumps are automatically started upon receipt of a SIAS
and all essential valves are aligned to their post-accident
positions.
During normal operation, both subsystems are required, and are independent to the degree necessary to assure the safe
operation and shutdown of the plant-assuming a single
failure. During shutdown, operation of the SRW System is the same as normal operation, except that the heat loads are
reduced. Additional information about the design and operation of the SRW System, along with a list of the
components served, is presented in Reference 1, Section 9.5.2.2. In the event of a LOCA, the SRW System
automatically realigns to isolate Turbine Building (non-
safety-related) loads creating two independent and redundant
safety-related subsystems. Service water flow to the spent
fuel pool (SFP) cooler and the blowdown heat exchanger is automatically isolated as required for the DBA. Each SRW subsystem will supply cooling water to a diesel generator
and two containment air coolers. However, the No. 11 SRW
subsystem only supplies two containment air coolers since
the No. 1A Diesel Generator is air cooled. Each SRW
subsystem is sufficiently sized to remove the maximum amount
of heat from the containment atmosphere while maintaining SRW System B 3.7.6 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.6-2 Revision 41 the SRW supply temperature to the diesel generator below its design limit.
APPLICABLE The design basis of the SRW System is for it to support a SAFETY ANALYSES 100% capacity containment cooling system (containment coolers) and to remove core decay heat 30 minutes following a design basis LOCA, as discussed in Reference 1, Section 14.20. This prevents the containment sump fluid
from increasing in temperature during the recirculation
phase following a LOCA and provides for a gradual reduction
in the temperature of this fluid as it is supplied to the
RCS by the safety injection pumps. The SRW System is
designed to perform its function with a single failure of
any active component, assuming the loss of offsite power.
The SRW System satisfies 10 CFR 50.36(c)(2)(ii), Criterion 3.
LCO Two SRW subsystems are required to be OPERABLE to provide the required redundancy to ensure that the system functions
to remove post-accident heat loads, assuming the worst
single active failure occurs coincident with the loss of
offsite power. Additionally, this system will also operate
assuming that worst case passive failure post-RAS.
An SRW subsystem is considered OPERABLE when: a. The associated pump and head tank are OPERABLE; and
- b. The associated piping, valves, heat exchanger, and instrumentation and controls required to perform the safety-related function are OPERABLE.
APPLICABILITY In MODEs 1, 2, 3, and 4, the SRW System is a normally operating system, which is required to support the
OPERABILITY of the equipment serviced by the SRW System and
required to be OPERABLE in these MODEs.
In MODEs 5 and 6, the OPERABILITY requirements of the SRW System are determined by the systems it supports.
SRW System B 3.7.6 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.6-3 Revision 5 ACTIONS A.1 and A.2 With one SRW heat exchanger inoperable, action must be taken to restore operable status within 7 days. Isolating flow to one associated containment cooling unit will reduce the DBA heat load of the affected SRW subsystem to within the capacity of one SRW heat exchanger, thus ensuring that the SRW temperatures can be maintained within their design limits. This will allow the associated diesel generator (except for 11 SRW which does not cool a diesel generator) to remain operable. In this Condition, the other OPERABLE SRW System is adequate to perform the containment heat removal function. However, the overall reliability is reduced because a single failure in the SRW System could result in loss of SRW containment heat removal function.
Required Action A.1 is modified by a Note. The Note indicates that the applicable Conditions of LCO 3.6.6 should be entered for an inoperable containment cooling train. The 7 day Completion Time is based on the redundant capabilities afforded by the OPERABLE subsystem, the Completion Time associated with an inoperable containment cooling unit (3.6.6), and the low probability of a DBA occurring during this time period. B.1 With one SRW subsystem inoperable, action must be taken to restore OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In this Condition, the remaining OPERABLE SRW System is adequate to perform the
heat removal function. However, the overall reliability is
reduced because a single failure in the SRW System could
result in loss of SRW function. Required Action B.1 is modified by a Note. The Note indicates that the applicable
Conditions of LCO 3.8.1, should be entered if the inoperable
SRW subsystem results in an inoperable diesel generator.
The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is based on the redundant
capabilities afforded by the OPERABLE subsystem, and the low probability of a DBA occurring during this time period.
C.1 and C.2 If the SRW subsystem cannot be restored to OPERABLE status within the associated Completion Time, the unit must be
placed in a MODE in which the LCO does not apply. To SRW System B 3.7.6 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.6-4 Revision 55 achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions
from full power conditions in an orderly manner and without challenging unit systems.
SURVEILLANCE SR 3.7.6.1 REQUIREMENTS Verifying the correct alignment for manual, power-operated, and automatic valves in the SRW flow path ensures that the
proper flow paths exist for SRW operation. This SR does not
apply to valves that are locked, sealed, or otherwise
secured in position, since they are verified to be in the
correct position prior to locking, sealing, or securing.
This SR also does not apply to valves that cannot be
inadvertently misaligned, such as check valves. This SR
does not require any testing or valve manipulation; rather, it involves verification that those valves capable of
potentially being mispositioned are in the correct position.
This SR is modified by a Note indicating that the isolation
of the SRW components or systems may render those components
inoperable but does not affect the OPERABILITY of the SRW
System.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.7.6.2 This SR verifies proper automatic operation of the SRW System valves on an actual or simulated actuation signal (SIAS or CSAS). The SRW System is a normally operating
system that cannot be fully actuated as part of normal
testing. This surveillance test is not required for valves
that are locked, sealed, or otherwise secured in the
required position under administrative controls. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SRW System B 3.7.6 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.6-5 Revision 55 SR 3.7.6.3 The SR verifies proper automatic operation of the SRW System pumps on an actual or simulated actuation signal (SIAS or
CSAS). The SRW System is a normally operating system that
cannot be fully actuated as part of the normal testing
during normal operation. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES 1. UFSAR
SW System B 3.7.7 B 3.7 PLANT SYSTEMS
B 3.7.7 Saltwater (SW) System
BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.7-1 Revision 5 BACKGROUND The SW System provides a heat sink for the removal of process and operating heat from safety-related components
during a DBA or transient. During normal operation or a
normal shutdown, the SW System also provides this function for various safety-related and non-safety-related components. The safety-related function is covered by this
LCO.
The SW System consists of two subsystems. Each subsystem contains one pump. A third pump, which is an installed
spare, can be aligned to either subsystem. The safety-
related function of each subsystem is to provide SW to two SRW heat exchangers, a CC heat exchanger, and an Emergency Core Cooling System (ECCS) pump room air cooler in order to
transfer heat from these systems to the bay. Seal water for
the non-safety-related circulating water pumps is supplied
by both or either subsystems. The SW pumps provide the
driving head to move SW from the intake structure, through
the system and back to the circulating water discharge
conduits. The system is designed such that each pump has
sufficient head and capacity to provide cooling water such
that 100% of the required heat load can be removed by either
subsystem.
During normal operation, both subsystems in each unit are in operation with one pump running on each header and a third
pump in standby. If needed, the standby pumps can be lined-
up to either supply header. The SW flow through the SRW and
CC heat exchangers is throttled to provide sufficient
cooling to the heat exchangers, while maintaining total
subsystem flow below a maximum value.
Additional information about the design and operation of the SW System, along with a list of the components served, is
presented in Reference 1. During an accident, the SW System
is required to remove the heat load from the SRW and ECCS pump room, and from the CC following an RAS.
SW System B 3.7.7 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.7-2 Revision 56 APPLICABLE The most limiting event for the SW System is a LOCA. SAFETY ANALYSES Operation of the SW System following a LOCA is separated into two phases, before the RAS and after the RAS. One
subsystem can satisfy cooling requirements of both phases.
After a LOCA but before a RAS, each subsystem will cool two SRW heat exchangers and an ECCS pump room air cooler (as required). There is no required flow to the CC heat exchangers. When an RAS occurs, flow is throttled to the CC
heat exchanger. Flow to each SRW heat exchanger is reduced
while the system remains capable of providing the required
flow to the ECCS pump room air coolers.
The SW System satisfies 10 CFR 50.36(c)(2)(ii), Criterion 3.
LCO Two SW subsystems are required to be OPERABLE to provide the required redundancy to ensure that the system functions to
remove post-accident heat loads, assuming the worst single
active failure occurs coincident with the loss of offsite
power. Additionally, this system will also operate assuming
the worst case passive failure post-RAS.
A SW subsystem is considered OPERABLE when: a. The associated pump is OPERABLE; and b. The associated piping, valves, heat exchangers, and instrumentation and controls required to perform the safety-related function are OPERABLE.
APPLICABILITY In MODEs 1, 2, 3, and 4, the SW System is a normally operating system, which is required to support the
OPERABILITY of the equipment serviced by the SW System and
required to be OPERABLE in these MODEs.
In MODEs 5 and 6, the OPERABILITY requirements of the SW System are determined by the systems it supports.
ACTIONS A.1 With one SW subsystem inoperable, action must be taken to restore OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In this Condition, the remaining OPERABLE SW subsystem is adequate to perform
the heat removal function. However, the overall reliability
is reduced because a single failure in the SW subsystem SW System B 3.7.7 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.7-3 Revision 2 could result in loss of SW System function. Required Action A.1 is modified by two Notes. The first Note
indicates that the applicable Conditions of LCO 3.8.1 should be entered if the inoperable SW subsystem results in an
inoperable emergency diesel generator. The second Note
indicates that the applicable Conditions and Required Actions of LCO 3.4.6 should be entered if an inoperable SW subsystem results in an inoperable SDC. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is based on the redundant capabilities
afforded by the OPERABLE train, and the low probability of a
DBA occurring during this time period.
B.1 and B.2 If the SW subsystems cannot be restored to OPERABLE status within the associated Completion Time, the unit must be
placed in a MODE in which the LCO does not apply. To
achieve this status, the unit must be placed in at least
MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions
from full power conditions in an orderly manner and without challenging unit systems.
SURVEILLANCE SR 3.7.7.1 REQUIREMENTS Verifying the correct alignment for manual, power-operated, and automatic valves in the SW System flow path ensures that
the proper flow paths exist for SW System operation. This
SR does not apply to valves that are locked, sealed, or otherwise secured in position, since they are verified to be in the correct position prior to locking, sealing, or
securing. This SR also does not apply to valves that cannot
be inadvertently misaligned, such as check valves. This
surveillance test does not require any testing or valve manipulation; rather, it involves verification that those
valves capable of potentially being mispositioned are in the
correct position. This SR is modified by a Note indicating
that the isolation of the SW System components or systems
may render those components inoperable but does not affect
the OPERABILITY of the SW System.
SW System B 3.7.7 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.7-4 Revision 56 The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.7.7.2 This SR verifies proper automatic operation of the SW System valves on an actual or simulated actuation signal (SIAS).
The SW System is a normally operating system that cannot be fully actuated as part of the normal testing. This
surveillance test is not required for valves that are
locked, sealed, or otherwise secured in the required
position under administrative controls. The Surveillance
Frequency is controlled under the Surveillance Frequency
Control Program. Note: There are currently no SW valves
with an Engineered Safety Feature Actuation System signal
since automatic system reconfiguration during a LOCA is not
required.
SR 3.7.7.3 This SR verifies proper automatic operation of the SW System pumps on an actual or simulated actuation signal (SIAS).
The SW System is a normally operating system that cannot be
fully actuated as part of the normal testing during normal
operation. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES 1. UFSAR, Section 9.5.2.3, "Saltwater System" CREVS B 3.7.8 B 3.7 PLANT SYSTEMS
B 3.7.8 Control Room Emergency Ventilation System (CREVS)
BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.8-1 Revision 42 BACKGROUND The CREVS provides a protected environment from which occupants can control the unit following an uncontrolled release of radioactivity, hazardous chemicals, or smoke.
The CREVS is a shared system providing protection for both Unit 1 and Unit 2.
The CREVS consists of two trains, including redundant
outside air intake ducts and redundant emergency
recirculation filter trains that recirculate and filter the
Control Room envelope (CRE) air and a CRE boundary that limits the inleakage of unfiltered air. The CREVS also has shared equipment, including an exhaust-to-atmosphere duct
containing redundant isolation valves and a normally closed
roof-mounted hatch, an exhaust-to-atmosphere duct from the
kitchen and toilet area of the Control Room containing a
single isolation valve, and common supply and return ducts
in both the standby and emergency recirculation portions of
the system. The shared equipment is considered to be a part
of each CREVS train. Each CREVS emergency recirculation
filter train consists of a prefilter, two high efficiency
particulate air (HEPA) filters for removal of aerosols, an
activated charcoal adsorber section for removal of elemental
and organic iodine and a fan. Ductwork, valves or dampers, doors, and barriers also form part of the system.
Instrumentation which actuates the system is addressed in
LCOs 3.3.4 and 3.3.8.
The CRE is the area within the confines of the CRE boundary that contains the spaces that Control Room occupants inhabit to control the Unit during normal and accident conditions.
This area encompasses the Control Room and may encompass non-critical areas to which frequent personnel access or continuous occupancy is not necessary in the event of an accident. The CRE is protected during normal operation, natural events, and accident conditions. The CRE boundary is the combination of walls, floor, roof, ducting, doors, penetrations, and equipment that physically form the CRE.
The OPERABILITY of the CRE boundary must be maintained to ensure that the inleakage of unfiltered air into the CRE will not exceed the inleakage assumed in the licensing basis CREVS B 3.7.8 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.8-2 Revision 42 analysis of DBA consequences to CRE occupants. The CRE and its boundary are defined in the Control Room Envelope Habitability Program.
The CREVS is an emergency system, parts of which may also
operate during normal unit operations in the standby mode of
operation. Actuation of the CREVS ensures the system is in
the emergency recirculation mode of operation, ensures the
unfiltered outside air intake and unfiltered exhaust-to-
atmosphere valves are closed, and aligns the system for
emergency recirculation of CRE air through the redundant trains of HEPA and charcoal filters. The prefilters remove
any large particles in the air and any entrained water
droplets present to prevent excessive loading of the HEPA
filters and charcoal adsorbers. A control room
recirculation signal (CRRS) initiates this filtered
ventilation of the air supply to the CRE.
The air recirculating through the CRE is continuously monitored by a radiation detector. Detector output above
the setpoint will cause actuation of the CREVS. The CREVS
operation in maintaining the Control Room habitable is
discussed in Reference 1, Section 9.8.2.3.
The redundant emergency recirculation filter train provides the required filtration should an excessive pressure drop
develop across the other filter train. A normally closed
hatch and double isolation valves are arranged in series to
prevent a breach of isolation from the outside atmosphere, except for the exhaust from the Control Room kitchen and
toilet areas. The CREVS is designed in accordance with
Seismic Category I requirements.
The CREVS is designed to maintain a habitable environment in the CRE for 30 days of continuous occupancy after a DBA without exceeding a 5 rem TEDE for the duration of the accident.
APPLICABLE The CREVS components are generally arranged in redundant SAFETY ANALYSES safety-related ventilation trains although some equipment is shared between trains.
The CREVS provides automatic airborne radiological protection for the CRE occupants, as demonstrated by the CRE CREVS B 3.7.8 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.8-3 Revision 42 occupant dose analyses for the most limiting design basis fission product release presented in Reference 1, Section 14.24.
The CREVS provides protection from smoke and hazardous chemicals to the CRE occupants. The analysis of hazardous chemical releases demonstrates that the toxicity limits are not exceeded in the CRE following a hazardous chemical release. The evaluation of a smoke challenge demonstrates that it will not result in the inability of the CRE occupants to control the reactor either from the Control Room or from the remote shutdown panels.
The CREVS also provides automatically actuated airborne radiological protection for the Control Room operations, for
the design basis fuel handling accident presented in
Reference 1, Section 14.18, the control element assembly
ejection event (Reference 1, Section 14.13, the main steam
line break (Reference 1, Section 14.14), the steam generator
tube rupture (Reference 1, Section 14.15), and the seized
rotor event (Reference 1, Section 14.16). The fuel handling
accident does not assume a single failure to occur.
The worst case single active failure of a component of the CREVS, assuming a loss of offsite power, does not impair the
ability of the system to perform its design function (except
for one valve in the shared duct between the Control Room
and the emergency recirculation filter trains).
The CREVS satisfies 10 CFR 50.36(c)(2)(ii), Criterion 3.
LCO The CREVS is required to be OPERABLE to ensure that the Control Room is isolated and at least one emergency
recirculation filter train is available, assuming a single active failure. Total system failure could result in exceeding a dose of 5 rem TEDE in the event of a large radioactive release.
The CREVS is considered OPERABLE when the individual components necessary to limit CRE occupant exposure are OPERABLE. For MODEs 1, 2, 3, and 4, redundancy is required
and CREVS is considered OPERABLE when: a. Both supply fans are OPERABLE; CREVS B 3.7.8 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.8-4 Revision 42 b. Both recirculation fans are OPERABLE; c. Both fans included in the emergency recirculation filter trains are OPERABLE; d. Both HEPA filters and charcoal adsorbers are not excessively restricting flow, and are capable of
performing their filtration functions; e. Ductwork, valves, and dampers are OPERABLE, such that air circulation can be maintained; and f. The Control Room outside air intake can be isolated for the emergency recirculation mode of operation, assuming
a single failure.
In order for the CREVS trains to be considered OPERABLE, the CRE boundary must be maintained such that the CRE occupant dose from a large radioactive release does not exceed the calculated dose in the licensing basis consequence analysis for DBAs, and that CRE occupants are protected from hazardous chemicals and smoke.
The LCO is modified by a Note which indicates that only one CREVS redundant component is required to be OPERABLE during
movement of irradiated fuel assemblies, when both units are
in MODEs 5 or 6, or defueled. Therefore, with both units in
other than MODEs 1, 2, 3, or 4, redundancy is not required
for movement of irradiated fuel assemblies and CREVS is
considered OPERABLE when: a. One supply fan is OPERABLE;
- b. One recirculation fan is OPERABLE;
- c. One fan included in the OPERABLE emergency recirculation filter train is OPERABLE; d. One train of two HEPA filters and one charcoal adsorber are not excessively restricting flow, and are capable
of performing their filtration functions; and e. Associated ductwork, valves, and dampers are OPERABLE, such that air circulation can be maintained and the
Control Room can be isolated for the emergency
recirculation mode.
When implementing the Note (since redundancy is not required), only one of the two isolation valves in each CREVS B 3.7.8 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.8-5 Revision 42 outside air intake duct is required, and only one of the two isolation valves in the exhaust to atmosphere duct is
required. However, the non-operating flow path must be
capable of providing isolation of the Control Room from the
outside atmosphere.
The LCO is modified by a second Note which indicates that only one CREVS train is required to be OPERABLE for the
movement of irradiated fuel assemblies. Therefore, redundancy is not required for movement of irradiated fuel
assemblies and only one CREVS train is required to be OPERABLE.
The LCO is modified by a third Note allowing the CRE boundary to be opened intermittently under administrative controls. This Note only applies to openings in the CRE boundary that can be rapidly restored to the design condition, such as doors, hatches, floor plugs, and access panels. For entry and exit through doors, the administrative control of the opening is performed by the person(s) entering or exiting the area. For other openings, these controls should be proceduralized and consist of stationing a dedicated individual at the opening who is in continuous communication with the operators in the CRE.
This individual will have a method to rapidly close the opening and to restore the CRE boundary to a condition equivalent to the design condition when the need for CRE isolation is indicated.
APPLICABILITY In MODEs 1, 2, 3, and 4, the CREVS must be OPERABLE to ensure that the CRE will remain habitable during and following a DBA.
During movement of irradiated fuel assemblies, the CREVS must be OPERABLE to cope with the release from a fuel
handling accident.
ACTIONS A.1 With one or more ducts with one Control Room outside air intake isolation valve inoperable in MODEs 1, 2, 3, or 4, the OPERABLE Control Room outside air intake valve in each
affected duct must be closed immediately. This places the CREVS B 3.7.8 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.8-6 Revision 42 OPERABLE Control Room outside air intake isolation valve in each affected duct in its safety function required position.
B.1 With the toilet area exhaust isolation valve inoperable, action must be taken to restore OPERABLE status within
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. In this Condition, the toilet area exhaust cannot
be isolated, therefore, the valve must be restored to
OPERABLE status. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period allows enough time to
repair the valve while limiting the time the toilet area is
open to the atmosphere. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time is based on the low probability of a DBA occurring during this time period.
C.1 With one exhaust to atmosphere isolation valve inoperable in MODEs 1, 2, 3, or 4, action must be taken to restore
OPERABLE status within seven days. In this Condition, the
remaining OPERABLE exhaust to atmosphere isolation valve is
adequate to isolate the Control Room. However, the overall
reliability is reduced because a single failure in the
OPERABLE exhaust to atmosphere isolation valve could result
in loss of exhaust to atmosphere isolation valve function.
The seven day Completion Time is based on the low
probability of a DBA occurring during this time period, and
the ability of the remaining exhaust to atmosphere isolation
valve to provide the required isolation capability.
D.1, D.2, and D.3 If the unfiltered inleakage of potentially contaminated air past the CRE boundary and into the CRE can result in CRE occupant radiological dose greater than the calculated dose of the licensing basis analyses of DBA consequences (allowed to be up to 5 rem TEDE), or inadequate protection of CRE occupants from hazardous chemicals or smoke, the CRE boundary is inoperable. Actions must be taken to restore an OPERABLE CRE boundary within 90 days.
During the period that the CRE boundary is considered inoperable, action must be initiated to implement mitigation actions to lessen the effect on CRE occupants from the potential hazards of a radiological or chemical event or a CREVS B 3.7.8 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.8-7 Revision 52 challenge from smoke. Required Action D.3 allows time to restore the CRE boundary to OPERABLE status provided
mitigating actions can ensure that the CRE remains within
the licensing basis habitability limits for the occupants
following an accident. Compensatory measures are discussed
in Reference 2. These compensatory measures may also be used as mitigating actions as required by Required
Action D.2. Temporary analytical methods may also be used
as compensatory measures to restore OPERABILITY. Actions
must be taken within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to verify that, in the event
of a DBA, the mitigating actions will ensure that CRE occupant radiological exposures will not exceed the calculated dose of the licensing basis analysis of DBA
consequences, and that CRE occupants are protected from
hazardous chemicals and smoke. These mitigating actions (i.e., actions that are taken to offset the consequences of
the inoperable CRE boundary) should be preplanned for
implementation upon entry into the condition, regardless of
whether entry is intentional or unintentional.
The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time is reasonable based on the
determination that the mitigating actions will ensure
protection of the CRE occupants within analyzed limits while
limiting the probability that CRE occupants will have to
implement protective measures that may adversely affect
their ability to control the reactor and maintain it in a
safe shutdown condition in the event of a DBA. In addition, the 90 day Completion Time is a reasonable time to diagnose, plan, and possibly repair and test most problems with the
CRE boundary.
E.1 With one CREVS train inoperable for reasons other than Conditions A, B, C, or D in MODEs 1, 2, 3, or 4, action must
be taken to restore OPERABLE status within seven days. In
this Condition, the remaining OPERABLE CREVS subsystem is
adequate to perform CRE occupant protection function.
However, the overall reliability is reduced because a
failure in the OPERABLE CREVS train could result in loss of CREVS function. The seven day Completion Time is based on the low probability of a DBA occurring during this time
period, and the ability of the remaining train to provide
the required capability.
CREVS B 3.7.8 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.8-8 Revision 52 F.1, F.2, and F.3 If both CREVS trains are inoperable in MODE 1, 2, 3, or 4 for reasons other than an inoperable Control Room boundary (i.e., Condition D), at least one CREVS train must be returned to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The Condition is modified by a Note stating it is not applicable if the second CREVS train is intentionally declared inoperable.
The Condition does not apply to voluntary removal of redundant systems or components from service. The Condition is only applicable if one train is inoperable for any reason and the second train is discovered to be inoperable, or if both trains are discovered to be inoperable at the same time. During the period that the CREVS trains are inoperable, action must be initiated to implement mitigating actions to lessen the effect on CRE occupants from potential hazards while both trains of CREVS are inoperable. In the event of a DBA, the mitigating actions will reduce the consequences of radiological exposures to the CRE occupants.
Specification 3.4.16, RCS Specific Activity, allows limited operation with the RCS activity significantly greater than the LCO limit. This presents a risk to the plant operator during an accident when all the CREVS trains are inoperable.
Therefore, it must be verified within one hour that LCO 3.4.16 is met. This Required Action does not require additional RCS sampling beyond that normally required by LCO 3.4.16.
At least one CREVS train must be returned to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The Completion Time is based on Reference 3 which demonstrated that the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time is acceptable based on the infrequent use of the Required Actions and the small incremental effect on plant risk.
G.1 Action G provides the actions to be taken when the Required Action and associated Completion Time of Condition B cannot
be met or with one or more CREVS trains inoperable due to an inoperable CRE boundary. It requires the immediate suspension of movement of irradiated fuel assemblies. This
places the unit in a condition that minimizes the accident
risk. This does not preclude the movement of fuel CREVS B 3.7.8 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.8-9 Revision 55 assemblies to a safe position. Since only one CREVS train must be OPERABLE for movement of irradiated fuel assemblies, the Required Action is applicable only to the required CREVS
train.
H.1 If both CREVS trains are inoperable for reasons other than Conditions A, B, C, or D, or if one or more ducts have two
outside air intake isolation valves inoperable, or if two
exhaust to atmosphere isolation valves are inoperable during
movement of irradiated fuel assemblies, the CREVS may not be capable of performing the intended function and the unit is in a condition outside the accident analyses. Therefore, movement of irradiated fuel must be suspended immediately.
This does not preclude the movement of fuel assemblies to a
safe condition.
I.1 and I.2 If the inoperable CREVs or Control Room boundary cannot be restored to OPERABLE status within the associated Completion
Time in MODE 1, 2, 3, or 4, the unit must be placed in a
mode that minimizes the accident risk. To achieve this
status the unit must be placed in at least MODE 3 within six
hours and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion
Times are reasonable, based on operating experience, to
reach the required unit conditions from full power
conditions in an orderly manner and without challenging unit systems. SURVEILLANCE SR 3.7.8.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. Since the environment and
normal operating conditions on this system are not severe, testing each required CREVS filter train once every month
provides an adequate check on this system.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
CREVS B 3.7.8 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.8-10 Revision 55 SR 3.7.8.2 This SR verifies that the required CREVS testing is performed in accordance with the Ventilation Filter Testing
Program (VFTP). The CREVS filter tests are in accordance
with portions of Reference 4. The VFTP includes testing
HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the
activated charcoal (general use and following specific
operations). Specific test Frequencies and additional
information are discussed in detail in the VFTP.
SR 3.7.8.3 This SR verifies each CREVS train starts and operates on an actual or simulated actuation signal (CRRS). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.7.8.4 This SR verifies the OPERABILITY of the CRE boundary by testing for unfiltered air inleakage past the CRE boundary
and into the CRE. The details of the testing are specified
in the Control Room Envelope Habitability Program.
The CRE is considered habitable when the radiological dose to the CRE occupants calculated in the licensing basis
analysis of DBA consequences is no more than 5 rem TEDE and
the CRE occupants are protected from hazardous chemicals and
smoke. This SR verifies that the unfiltered air inleakage
into the CRE is no greater than the flow rate assumed in the
licensing basis analysis of DBA consequences. When
unfiltered air inleakage is greater than the assumed flow
rate, Condition E must be entered. Options for restoring
the CRE boundary to OPERABLE status include changing the
licensing basis DBA consequences analysis, repairing the CRE
boundary, or a combination of these actions. Depending upon
the nature of the problem and the corrective action, a full
scope inleakage test may not be necessary to establish that the CRE boundary has been restored to OPERABLE status.
CREVS B 3.7.8 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.8-11 Revision 52 REFERENCES 1. UFSAR 2. Regulatory Guide 1.196, Revision 0, "Control Room Habitability at Light-Water Nuclear Power Reactors," May 2003 3. WCAP-16125-NP-A, "Justification for Risk-Informed Modifications to Selected Technical Specifications for Conditions Leading to Exigent Plant Shutdown,"
Revision 2, August 2010 4. Regulatory Guide 1.52, Revision 2, "Design, Testing, and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants," March 1978
CRETS B 3.7.9 B 3.7 PLANT SYSTEMS
B 3.7.9 Control Room Emergency Temperature System (CRETS)
BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.9-1 Revision 2 BACKGROUND The CRETS provides temperature control for the Control Room following isolation of the Control Room. The CRETS is a shared system which is supported by the CREVS, since the CREVS must be operating in the emergency recirculation mode for CRETS to perform its safety function.
The CRETS consists of two independent, redundant trains that provide cooling of recirculated Control Room air. Each train consists of cooling coils, instrumentation, and
controls to provide for Control Room temperature control.
The CRETS is a subsystem providing air temperature control
for the Control Room.
The CRETS is an emergency system, parts of which may also operate during normal unit operations in the standby mode of
operation. A single train will provide the required
temperature control to maintain the Control Room below 104°F. The CRETS operation to maintain the Control Room temperature is discussed in Reference 1.
APPLICABLE The design basis of the CRETS is to maintain temperature SAFETY ANALYSES of the Control Room environment throughout 30 days of continuous occupancy.
The CRETS components are arranged in redundant safety-related trains. During emergency operation, the CRETS
maintains the temperature below 104°F. A single active
failure of a component of the CRETS, assuming a loss of offsite power, does not impair the ability of the system to perform its design function. Redundant detectors and
controls are provided for Control Room temperature control.
The CRETS is designed in accordance with Seismic Category I
requirements. The CRETS is capable of removing sensible and
latent heat loads from the Control Room, considering equipment heat loads and personnel occupancy requirements, to ensure equipment OPERABILITY.
The CRETS satisfies 10 CFR 50.36(c)(2)(ii), Criterion 3.
CRETS B 3.7.9 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.9-2 Revision 31 LCO Two independent and redundant trains of the CRETS are required to be OPERABLE to ensure that at least one is
available, assuming a single failure disables the other
train following isolation of the Control Room. Total system
failure could result in the equipment operating temperature
exceeding limits in the event of an accident requiring isolation of the Control Room.
The CRETS is considered OPERABLE when the individual components that are necessary to maintain the Control Room
temperature are OPERABLE. The required components include
the cooling coils and associated temperature control
instrumentation. In addition, the CRETS must be OPERABLE to
the extent that air circulation can be maintained.
For MODEs 1, 2, 3, and 4, redundancy is required and both trains must be OPERABLE. The LCO is modified by a Note
which indicates that only one CRETS train is required to be
OPERABLE for the movement of irradiated fuel assemblies.
Therefore, redundancy is not required for movement of
irradiated fuel assemblies and only one CRETS train is required to be OPERABLE.
APPLICABILITY In MODEs 1, 2, 3, and 4, and during movement of irradiated fuel assemblies, the CRETS must be OPERABLE to ensure that
the Control Room temperature will not exceed equipment
OPERABILITY requirements following isolation of the Control
Room. ACTIONS A.1 With one CRETS train inoperable in MODEs 1, 2, 3, or 4, action must be taken to restore OPERABLE status within
30 days. In this Condition, the remaining OPERABLE CRETS
train is adequate to maintain the Control Room temperature
within limits. The 30 day Completion Time is reasonable, based on the low probability of an event occurring requiring
Control Room isolation, consideration that the remaining
train can provide the required capabilities, and the
alternate safety or non-safety-related cooling means that
are available.
CRETS B 3.7.9 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.9-3 Revision 55 B.1 and B.2 If the Required Actions and associated Completion Times of Condition A are not met in MODEs 1, 2, 3, or 4, the unit
must be placed in a MODE that minimizes the accident risk.
To achieve this status, the unit must be placed in at least
MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full
power conditions in an orderly manner and without
challenging unit systems.
C.1 If both CRETS trains are inoperable in MODEs 1, 2, 3, or 4, or during movement of irradiated fuel assemblies, the CRETS
may not be capable of performing the intended function and
the unit is in a condition outside the accident analysis.
Therefore, LCO 3.0.3 must be entered immediately and
movement of irradiated fuel must be suspended immediately.
This does not preclude the movement of fuel assemblies to a safe condition.
SURVEILLANCE SR 3.7.9.1 REQUIREMENTS This SR verifies each required CRETS train has the capability to maintain Control Room temperature 104°F for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in the recirculation mode. During this test, the backup Control Room air conditioner is to be de-
energized. This SR consists of a combination of testing.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES 1. UFSAR, Section 9.8.2.3, "Auxiliary Building Ventilating Systems" SFPEVS B 3.7.11 B 3.7 PLANT SYSTEMS
B 3.7.11 Spent Fuel Pool Exhaust Ventilation System (SFPEVS)
BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.11-1 Revision 41 BACKGROUND The SFPEVS exhausts airborne radioactive particulates and gases from the area of the fuel pool into the plant ventilation stack following a fuel handling accident involving recently irradiated fuel.
The SFPEVS consists of two independent, redundant exhaust fans. Ductwork, valves or dampers, and instrumentation also form part of the system. The SFPEVS is supplied power by
one non-safety-related power supply.
The SFPEVS is operated during normal unit operations. When movement of the air is required (i.e., during movement of recently irradiated fuel assemblies in the Auxiliary
Building), normal air discharges from the fuel handling area
in the Auxiliary Building.
The SFPEVS is discussed in Reference 1, Sections 9.8.2.3 and 14.18, because it may be used for normal, as well as post-accident ventilation.
APPLICABLE The SFPEVS is designed to mitigate the consequences of a SAFETY ANALYSES fuel handling accident involving handling recently irradiated fuel (i.e., fuel that has occupied part of a
critical reactor core within the previous 55 days), in which all rods in the fuel assembly are assumed to be damaged.
The analysis of the fuel handling accident is given in
Reference 1, Section 14.18. The DBA analysis of the fuel
handling accident assumes that the SFPEVS is functional and exhausts airborne radioactive particulates and gases from the fuel pool area into the plant ventilation stack. The analysis follows the guidance provided in Reference 2.
The SFPEVS satisfies 10 CFR 50.36(c)(2)(ii), Criterion 3.
LCO Two exhaust fans and other equipment listed in the Background Section are required to be OPERABLE and in
operation.
The SFPEVS is considered OPERABLE when the individual components necessary to direct exhaust into the ventilation SFPEVS B 3.7.11 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.11-2 Revision 41 stack are OPERABLE. The SFPEVS is considered OPERABLE when its associated: a. Fans are OPERABLE; and b. Ductwork, valves, and dampers are OPERABLE, and air circulation can be maintained.
The SFPEVS is considered in operation when an OPERABLE exhaust fan is in operation.
APPLICABILITY During movement of recently irradiated fuel assemblies in the Auxiliary Building, the SFPEVS is required to be
OPERABLE and in operation to mitigate the consequences of a
fuel handling accident involving handling recently
irradiated fuel by minimizing the atmospheric dispersion to the Control Room. Due to radioactive decay, the SFPEVS is only required to mitigate fuel handling accidents involving
handling recently irradiated fuel (i.e., fuel that has
occupied part of a critical reactor core within the previous 55 days).
ACTIONS A.1 and A.2 When one SFPEVS exhaust fan is inoperable, action must be taken to verify an OPERABLE SFPEVS train is in operation, or
movement of recently irradiated fuel assemblies in the
Auxiliary Building must be suspended. One OPERABLE SFPEVS train consists of one OPERABLE exhaust fan. This ensures the proper equipment is operating for the Applicable Safety Analysis.
B.1 When there is no OPERABLE SFPEVS train or there is no OPERABLE SFPEVS train in operation during movement of
recently irradiated fuel assemblies in the Auxiliary
Building, action must be taken to place the unit in a
condition in which the LCO does not apply. This Action
involves immediately suspending movement of recently
irradiated fuel assemblies in the Auxiliary Building. This does not preclude the movement of fuel to a safe position.
SFPEVS B 3.7.11 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.11-3 Revision 55 SURVEILLANCE SR 3.7.11.1 REQUIREMENTS The SR requires verification that the SFPEVS is in operation. Verification includes verifying that one exhaust
fan is operating and discharging into the ventilation stack.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.7.11.2 Deleted.
SR 3.7.11.3 This SR verifies the integrity of the spent fuel storage pool area. The ability of the spent fuel storage pool area
to maintain negative pressure with respect to potentially
uncontaminated adjacent areas is periodically tested to
verify proper function of the SFPEVS. During operation, the
spent fuel storage pool area is designed to maintain a
slight negative pressure in the spent fuel storage pool
area, with respect to adjacent areas, to ensure that
exhausted air is directed to the ventilation stack.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES 1. UFSAR 2. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000
PREVS B 3.7.12 B 3.7 PLANT SYSTEMS
B 3.7.12 Penetration Room Exhaust Ventilation System (PREVS)
BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.12-1 Revision 52 BACKGROUND The PREVS filters air from the penetration room.
The PREVS consists of two independent and redundant trains.
Each train consists of a prefilter, a HEPA filter, an activated charcoal adsorber section for removal of gaseous activity (principally iodines), and a fan. Ductwork, valves
or dampers, and instrumentation also form part of the
system. The system initiates filtered ventilation following
receipt of a containment isolation actuation signal.
The PREVS is a standby system, which may also operate during normal unit operations. During emergency operations, the
PREVS dampers are realigned, and fans are started to
initiate filtration. Upon receipt of the actuating
Engineered Safety Feature Actuation System signal(s), normal
air discharges from the penetration room, and the stream of
ventilation air discharges through the system filter trains.
The prefilters remove any large particles in the air to
prevent excessive loading of the HEPA filters and charcoal
adsorbers.
The PREVS is discussed in Reference 1, Section 6.6.2, as it may be used for normal, as well as post-accident, atmospheric cleanup functions.
APPLICABLE The design basis of the PREVS is established by the Maximum SAFETY ANALYSES Hypothetical Accident. The system is credited with filtering the radioactive material released through the containment vent when the line is open. Also commensurate with the guidance in Reference 2, a conservative bypass fraction from the Containment to the penetration rooms is
assumed. Following a LOCA, the containment isolation signal
will start both of the fans associated with the PREVS, filtering the exhaust through the HEPA and charcoal filters, and directing the exhaust into the ventilation stack. The
analysis of the effects and consequences of a Maximum
Hypothetical Accident are presented in Reference 1, Section 14.24 and follows the guidance presented in
Reference 3.
PREVS B 3.7.12 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.12-2 Revision 41 As a layer of defense, the Penetration Room Exhaust Ventilation System also provides filtered ventilation of radioactive materials leaking from ECCS equipment within the
penetration room following an accident, however, credit for
this feature was not assumed in the accident analysis (Reference 1, Section 14.24).
The PREVS satisfies 10 CFR 50.36(c)(2)(ii), Criterion 3.
LCO Two independent and redundant trains of the PREVS are required to be OPERABLE to ensure that at least one train is
available, assuming there is a single failure disabling the
other train coincident with a loss of offsite power.
The PREVS is considered OPERABLE when the individual components necessary to control radioactive releases are
OPERABLE in both trains. A PREVS train is considered
OPERABLE when its associated: a. Fan is OPERABLE;
- b. High efficiency particulate air filter and charcoal adsorber are not excessively restricting flow, and are
capable of performing the filtration functions; and c. Ductwork, valves, and dampers are OPERABLE, and circulation can be maintained.
APPLICABILITY In MODEs 1, 2, and 3, the PREVS is required to be OPERABLE to mitigate the potential radioactive material release from
a Maximum Hypothetical Accident.
In MODEs 4, 5, and 6, the PREVS is not required to be OPERABLE, since the RCS temperature and pressure are low and
there is insufficient energy to result in the conditions assumed in the accident analysis.
ACTIONS A.1 With one PREVS train inoperable, action must be taken to restore OPERABLE status within seven days. During this time
period, the remaining OPERABLE train is adequate to perform
the PREVS function. The seven day Completion Time is
reasonable based on the low probability of a DBA occurring PREVS B 3.7.12 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.12-3 Revision 52 during this time period, and the consideration that the remaining train can provide the required capability.
B.1 and B.2 With two PREVS trains inoperable, action must be taken to restore at least one PREVS train to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The Condition is modified by a Note stating it is not applicable if the second PREVS train is intentionally declared inoperable. The Condition does not apply to voluntary removal of redundant systems or components from service. The Condition is only applicable if one train is inoperable for any reason and the second train is discovered to be inoperable, or if both trains are discovered to be inoperable at the same time. In addition, at least one train of containment spray must be verified to be OPERABLE within one hour. In the event of an accident, containment spray reduces the potential radioactive release from the containment, which reduces the consequences of the inoperable PREVS trains. The Completion Time is based on Reference 4 which demonstrated that the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time is acceptable based on the infrequent use of the Required Actions and the small incremental effect on plant risk.
C.1 and C.2 If the inoperable train cannot be restored to OPERABLE status within the associated Completion Time, the unit must
be placed in a MODE in which the LCO does not apply. To
achieve this status, the unit must be placed in at least
MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The
allowed Completion Times are reasonable, based on operating
experience, to reach the required unit conditions from full
power conditions in an orderly manner and without challenging unit systems.
SURVEILLANCE SR 3.7.12.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. As the environment and normal
operating conditions on this system are not severe, testing
each train once every month provides an adequate check on
this system.
PREVS B 3.7.12 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.12-4 Revision 55 The test is performed by initiating the system from the Control Room, ensuring flow through the HEPA filter and
charcoal adsorber train, and verifying this system operates for 15 minutes. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.7.12.2 This SR verifies the performance of PREVS filter testing in accordance with the VFTP. The PREVS filter tests are in
accordance with portions of Reference 5. The VFTP includes
testing the performance of the HEPA filter, charcoal
adsorber efficiency, minimum system flow rate, and the
physical properties of the activated charcoal (general use
and following specific operations). Specific test
frequencies and additional information are discussed in
detail in the VFTP.
SR 3.7.12.3 This SR verifies that each PREVS train starts and operates on an actual or simulated actuation signal (Containment
Isolation Signal). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES 1. UFSAR 2. Regulatory Guide 1.194, Atmospheric Relative Concentrations for Control Room Radiological
Habitability Assessments at Nuclear Power Plants, June
2003 3. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000 4. WCAP-16125-NP-A, "Justification for Risk-Informed Modifications to Selected Technical Specifications for
Conditions Leading to Exigent Plant Shutdown,"
Revision 2, August 2010 PREVS B 3.7.12 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.12-5 Revision 55 5. Regulatory Guide 1.52, Revision 2, "Design, Testing, and Maintenance Criteria for Post Accident Engineered-
Safety-Feature Atmosphere Cleanup System Air Filtration
and Adsorption Units of Light-Water-Cooled Nuclear Power Plants," March 1978
SFP Water Level B 3.7.13 B 3.7 PLANT SYSTEMS
B 3.7.13 Spent Fuel Pool (SFP) Water Level
BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.13-1 Revision 41 BACKGROUND The minimum water level in the SFP meets the assumptions of iodine decontamination factors following a fuel handling
accident. The specified water level shields and minimizes
the general area dose when the storage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel.
A general description of the SFP design is given in Reference 1, Section 9.7.2, and the SFP Cooling and Cleanup
System is given in Reference 1, Section 9.4.1. The
assumptions of the fuel handling accident are given in Reference 1, Section 14.18.
APPLICABLE Per Reference 2, the Fuel Handling Accident (FHA) analysis SAFETY ANALYSES may assume a total iodine decontamination factor of 200 based on a minimum water depth of 23 feet. The minimum water level requirement ensures that sufficient water depth
is available to remove 99.5% of gap activity, which is comprised of 16% I-131 and 10% of all other iodine isotopes released from the rupture of an irradiated fuel assembly.
The Technical Specifications requirement of 21.5 feet of water above fuel assemblies seated in the SFP storage racks
is sufficient to preserve the required 23 feet of water
because an FHA was assumed to occur as a fuel assembly
strikes the bottom of the SFP.
When assemblies are placed on rack spacers with their upper end fittings removed, an FHA caused by a dropped heavy object would result in a lower decontamination factor based
on reduced water coverage. A revised decontamination factor
of 120 for an FHA during reconstitution or inspection with 20.4 feet of water between the top of the pin and the
surface of the water was computed for an assembly placed on a 20.5 inch rack spacer with its upper end fitting removed.
Note that this is very conservative, since normal level
control will result in at least 21.5 feet of water above
exposed fuel pins. This results in a 99.17% removal rate.
SFP Water Level B 3.7.13 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.13-2 Revision 55 The SFP water level satisfies 10 CFR 50.36(c)(2)(ii), Criteria 2 and 3.
LCO The specified water level preserves the assumptions of the fuel handling accident analysis (Reference 1, Section 14.18). As such, it is the minimum required for fuel storage, reconstitution, and movement within the fuel
storage pool.
APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the SFP since the potential for a release of fission products exists.
ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.
When the initial conditions for an accident cannot be met, steps should be taken to preclude the accident from
occurring. When the SFP water level is lower than the
required level, the movement of irradiated fuel assemblies
in the SFP is immediately suspended. This effectively
precludes a spent fuel handling accident from occurring.
This does not preclude moving a fuel assembly to a safe
position.
If moving irradiated fuel assemblies while in MODEs 5 or 6, LCO 3.0.3 would not specify any action. If moving
irradiated fuel assemblies while in MODEs 1, 2, 3, and 4, the fuel movement is independent of reactor operations.
Therefore, in either case, inability to suspend movement of
irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.
SURVEILLANCE SR 3.7.13.1 REQUIREMENTS This SR verifies sufficient SFP water is available in the event of a fuel handling accident. The water level in the
SFP must be checked periodically. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SFP Water Level B 3.7.13 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.13-3 Revision 41 During refueling operations, the level in the SFP is normally at equilibrium with that of the refueling canal.
REFERENCES 1. UFSAR 2. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000
Secondary Specific Activity B 3.7.14 B 3.7 PLANT SYSTEMS
B 3.7.14 Secondary Specific Activity
BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.14-1 Revision 41 BACKGROUND Activity in the secondary coolant results from steam generator tube outleakage from the RCS. Under steady state
conditions, the activity is primarily iodines with
relatively short half lives, and thus is an indication of current conditions. During transients, DOSE EQUIVALENT I-131 spikes have been observed as well as
increased releases of some noble gases. Other fission
product isotopes, as well as activated corrosion products in
lesser amounts, may also be found in the secondary coolant.
A limit on secondary coolant specific activity during power operation minimizes releases to the environment because of
normal operation, anticipated operational occurrences, and
accidents.
This limit is lower than the activity value that might be expected from a 100 gallons per day tube leak (LCO 3.4.13) of primary coolant at the limit of 0.5
µCi/gm (LCO 3.4.15).
The main SLB is assumed to result in the release of the noble gas and iodine activity contained in the steam
generator inventory, the feedwater, and reactor coolant
LEAKAGE via flashing directly to the environment through the main steam gooseneck.
APPLICABLE The accident analysis of the main SLB, as discussed in SAFETY ANALYSES Reference 1, assumes the initial secondary coolant specific activity to have a radioactive isotope concentration of 0.10 µCi/gm DOSE EQUIVALENT I-131. This secondary activity, together with the Technical Specification primary system activity, and failed fuel activity, is used in the analysis for determining the radiological consequences of the postulated accident. The accident analysis shows that the radiological consequences of a main SLB do not exceed
the acceptance criteria given in References 1 and 2.
With the loss of offsite power post-main SLB, the remaining steam generator is available for core decay heat dissipation
by venting steam to the atmosphere through MSSVs and ADVs.
The AFW System supplies the necessary makeup to the steam
generator. Venting continues until the reactor coolant Secondary Specific Activity B 3.7.14 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.14-2 Revision 41 temperature and pressure have decreased sufficiently for the SDC System to complete the cooldown.
Other accidents or transients, such as a steam generator tube rupture, a seized rotor event, and a control element assembly ejection event, involve a partial release of the secondary activity via steam release to the atmosphere via the ADVs and MSSVs. These releases contribute to the offsite and Control Room doses listed in Reference 1, Section 14. These accident analyses show that the radiological consequences of a DBA do not exceed the acceptance criteria given in References 1 and 2.
Secondary specific activity limits satisfy 10 CFR 50.36(c)(2)(ii), Criterion 2.
LCO As indicated in the Applicable Safety Analyses, the specific activity limit in the secondary coolant system of 0.10 µCi/gm DOSE EQUIVALENT I-131 limits the radiological consequences of a DBA to the acceptance criteria given in Reference 1.
Monitoring the specific activity of the secondary coolant ensures that when secondary specific activity limits are
exceeded, appropriate actions are taken in a timely manner
to place the unit in an operational MODE that would minimize the radiological consequences of a DBA.
APPLICABILITY In MODEs 1, 2, 3, and 4, the limits on secondary specific activity apply due to the potential for secondary steam
releases to the atmosphere.
In MODEs 5 and 6, the steam generators are not being used for heat removal. Both the RCS and steam generators are
depressurized, and primary to secondary LEAKAGE is minimal.
Therefore, monitoring of secondary specific activity is not required.
ACTIONS A.1 and A.2 DOSE EQUIVALENT I-131 exceeding the allowable value in the secondary coolant, is an indication of a problem in the RCS, and contributes to increased post-accident doses. If Secondary Specific Activity B 3.7.14 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.14-3 Revision 55 secondary specific activity cannot be restored to within limits in the associated Completion Time, the unit must be
placed in a MODE in which the LCO does not apply. To
achieve this status, the unit must be placed in at least
MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The
allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
SURVEILLANCE SR 3.7.14.1 REQUIREMENTS This SR ensures that the secondary specific activity is within the limits of the accident analysis. A gamma isotope
analysis of the secondary coolant, which determines DOSE
EQUIVALENT I-131, confirms the validity of the safety
analysis assumptions as to the source terms in post-accident
releases. It also serves to identify and trend any unusual
isotopic concentrations that might indicate changes in
reactor coolant activity or LEAKAGE. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES 1. UFSAR, Chapter 14, "Safety Analysis" 2. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000
MFIVs B 3.7.15 B 3.7 PLANT SYSTEMS
B 3.7.15 Main Feedwater Isolation Valves (MFIVs)
BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.15-1 Revision 2 BACKGROUND The MFIVs isolate MFW flow to the secondary side of the steam generators following a HELB. The consequences of HELBs occurring in the main steam lines or in the MFW lines downstream of the MFIVs will be mitigated by their closure.
Closure of the MFIVs effectively terminates the addition of feedwater to an affected steam generator, limiting the mass
and energy release for SLBs /or feedwater line breaks (FWLBs) inside the Containment Structure upstream of the reverse flow check valve, and reducing the cooldown effects
for SLBs.
The MFIVs isolate the non-safety-related portions from the safety-related portion of the system. In the event of a secondary side pipe rupture inside the Containment Structure upstream of the reverse flow check valve, the valves limit
the quantity of high energy fluid that enters the Containment Structure through the break.
One MFIV is located on each MFW line, outside, but close to, the Containment Structure. The MFIVs are located so that AFW may be supplied to a steam generator following MFIV
closure. The piping volume from the valve to the steam
generator must be accounted for in calculating mass and
energy releases.
The MFIVs close on receipt of a steam generator isolation signal generated by low steam generator pressure. The steam generator isolation signal also actuates the MSIVs to close.
The MFIVs may also be actuated manually. In addition, the MFIVs reverse flow check valve inside the Containment Structure is available to isolate the feedwater line penetrating the Containment Structure, and to ensure that the consequences of events do not exceed the capacity of the
Containment Cooling System.
A description of the MFIVs operation on receipt of an steam generator isolation signal is found in Reference 1.
MFIVs B 3.7.15 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.15-2 Revision 13 APPLICABLE The design basis of the MFIVs is established by the analysis SAFETY ANALYSES for the large SLB. It is also influenced by the accident analysis for the large FWLB.
Failure of an MFIV to close following an SLB or FWLB can result in additional mass and energy to the steam generator's contributing to cooldown. This failure also results in additional mass and energy releases following an
SLB or FWLB event.
The MFIVs satisfy 10 CFR 50.36(c)(2)(ii), Criterion 3.
LCO This LCO ensures that the MFIVs will isolate MFW flow to the steam generators. Following an FWLB or SLB, these valves
will also isolate the non-safety-related portions from the
safety-related portions of the system. This LCO requires
that one MFIV in each feedwater line be OPERABLE. The MFIVs
are considered OPERABLE when the isolation times are within
limits, and are closed on an isolation actuation signal.
Failure to meet the LCO requirements can result in additional mass and energy being released to the Containment
Structure following an SLB or FWLB inside the Containment
Structure. Failure to meet the LCO can also add additional
mass and energy to the steam generators contributing to cooldown.
APPLICABILITY The MFIVs must be OPERABLE whenever there is significant mass and energy in the RCS and steam generators.
In MODEs 1, 2, and 3, the MFIVs are required to be OPERABLE in order to limit the amount of available fluid that could
be added to the Containment Structure in the case of a
secondary system pipe break inside the Containment
Structure.
In MODEs 4, 5, and 6, steam generator energy is low.
MFIVs B 3.7.15 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.15-3 Revision 61 ACTIONS The ACTIONS table is modified by a Note indicating that separate Condition entry is allowed for each valve.
A.1 With one MFIV inoperable, action must be taken to restore the valve to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time takes into account the isolation capability afforded by the MFW regulating valves, and
tripping of the MFW pumps, and the low probability of an
event occurring during this time period that would require
isolation of the MFW flow paths.
B.1 and B.2 If the MFIVs cannot be restored to OPERABLE status in the associated Completion Time, the unit must be placed in a
MODE in which the LCO does not apply. To achieve this
status, the unit must be placed in at least MODE 3 within
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed
Completion Times are reasonable, based on operating
experience, to reach the required unit conditions from full
power conditions in an orderly manner and without challenging unit systems.
SURVEILLANCE SR 3.7.15.1 REQUIREMENTS This SR ensures the closure time for each MFIV is 65 seconds by manual isolation. The MFIV closure time is assumed in the accident and containment analyses.
The Frequency is in accordance with the INSERVICE TESTING PROGRAM. The MFIVs are tested during each refueling outage in accordance with Reference 2, and sometimes during other cold shutdown periods. The Frequency demonstrates the valve
closure time at least once per refueling cycle. Operating
experience has shown that these components usually pass the surveillance test when performed.
MFIVs B 3.7.15 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.15-4 Revision 38 REFERENCES 1. UFSAR, Section 14.4.2, "Sequence of Events" 2. ASME Code for Operation and Maintenance of Nuclear Power Plants
SFP Boron Concentration B 3.7.16 B 3.7 PLANT SYSTEMS
B 3.7.16 Spent Fuel Pool (SFP) Boron Concentration
BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.16-1 Revision 23 BACKGROUND Fuel assemblies are stored in the spent fuel racks in accordance with criteria based on 10 CFR 50.68. If credit is taken for soluble boron, the k-effective of the spent
fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95%
probability, 95% confidence level, if flooded with borated
water, and the k-effective must remain below 1.0 (subcritical) at a 95% probability, 95% confidence level, if
flooded with unborated water. In addition, the maximum
nominal U-235 enrichment of the fresh fuel assemblies is limited to 5.0 weight percent.
APPLICABLE The criticality analyses were done such that the criteria of SAFETY ANALYSES 10 CFR 50.68 are met. Boron dilution events are credible, postulated accidents, when credit for soluble boron is
taken. The minimum SFP boron concentration in this
Technical Specification supports the initial boron concentration assumption in the dilution calculations.
For other non-dilution accident scenarios, the double contingency principle of ANSI N 16.1-1975 requires two
unlikely, independent concurrent events to produce a
criticality accident and thus allows credit for the nominal
soluble boron concentration, as defined in LCO 3.7.16.
The concentration of dissolved boron in the SFPs satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO The specified concentration of dissolved boron in the SFP preserves the assumptions used in the analyses of the
potential accident scenarios described above. This
concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the SFPs.
APPLICABILITY This LCO applies whenever fuel assemblies are stored in the SFPs.
SFP Boron Concentration B 3.7.16 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.16-2 Revision 55 ACTIONS A.1 and A.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel
assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify
any action. If moving irradiated fuel assemblies while in
MODE 1, 2, 3, or 4, the fuel movement is independent of
reactor operation. Therefore, inability to suspend movement
of fuel assemblies is not a sufficient reason to require a
reactor shutdown.
When the concentration of boron in the SFPs is less than
required, immediate action must be taken to preclude an
accident from happening or to mitigate the consequences of
an accident in progress. This is most efficiently achieved
by immediately suspending the movement of fuel assemblies.
This does not preclude the movement of fuel assemblies to a safe position. In addition, action must be immediately
initiated to restore boron concentration to within limits.
SURVEILLANCE SR 3.7.16.1 REQUIREMENTS This SR verifies that the concentration of boron in the SFPs is within the required limit. As long as this SR is met, the analyzed incidents are fully addressed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES None
SFP Storage B 3.7.17 B 3.7 PLANT SYSTEMS B 3.7.17 Spent Fuel Pool (SFP) Storage BASES CALVERT CLIFFS - UNIT 2 B 3.7.17-1 Revision 23 BACKGROUND This Technical Specification applies to the Unit 2 SFP only.
The spent fuel storage facility was originally designed to store either new (non-irradiated) nuclear fuel assemblies or burned (irradiated) fuel assemblies in a vertical configuration underwater, assuming credit for Boraflex poison sheets but assuming no credit for soluble boron or burnup. The spent fuel storage cells are installed in parallel rows with center-to-center spacing of 10 3/32 inches and with Boraflex sheets between adjacent assemblies. This spacing was sufficient to maintain keff 0.95 for spent fuel of enrichments up to 4.52 wt% for standard fuel design and up to 4.30 wt% for Value Added Pellet (VAP) fuel design.
The burnup and enrichment requirements of LCO 3.7.17(a) ensures that the multiplication factor (keff) for the rack in the SFP is less than the 10 CFR 50.68 regulatory limit with the VAP fuel design, ranging in enrichment from 2.0 to 5.0 wt%, with burnup credit, with partial credit for soluble boron, but without Boraflex credit. The soluble boron credit will be limited to 350 ppm including all biases and uncertainties. For fuel assemblies which do not satisfy the burnup and enrichment requirements of LCO 3.7.17(a), the fuel assemblies may be stored in the Unit 2 SFP if surrounded on all four adjacent faces by empty rack cells or other non-reactive materials per LCO 3.7.17(b).
APPLICABLE The Unit 2 spent fuel storage facility is designed to SAFETY ANALYSES conform to the requirements of 10 CFR 50.68 by use of adequate spacing, soluble boron credit, and burnup credit.
The SFP storage satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO The restrictions on the placement of fuel assemblies within the Unit 2 SFP are in accordance with Figure 3.7.17-1 and ensure that the Unit 2 SFP meets the requirements of 10 CFR 50.68. The restrictions are consistent with the criticality safety analysis performed for the Unit 2 SFP. Fuel assemblies not meting the criteria of Figure 3.7.17-1 may be SFP Storage B 3.7.17 BASES CALVERT CLIFFS - UNIT 2 B 3.7.17-2 Revision 23 stored in the Unit 2 SFP in a checkboard pattern in accordance with LCO 3.7.17(b).
APPLICABILITY This LCO applies whenever any fuel assembly is stored in the Unit 2 SFP.
ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply. If moving fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation. Therefore, in either case, inability to move fuel assemblies is not a sufficient reason to require a reactor shutdown.
When the configuration of fuel assemblies stored in Unit 2 SFP is not in accordance with Figure 3.7.17-1 or LCO 3.7.17(b), immediate action must be taken to make the necessary fuel assembly movement(s) to bring the fuel assembly configuration into compliance with Figure 3.7.17-1 or LCO 3.7.17(b).
SURVEILLANCE SR 3.7.17.1 REQUIREMENTS This SR verifies by administrative means that the initial enrichment and burnup of the fuel assembly is in accordance with Figure 3.7.17-1 for LCO 3.7.17(a). This Surveillance Requirement does not address fuel assemblies stored in the Unit 2 SFP in accordance with LCO 3.7.17(b). This will ensure compliance with Specification 4.3.1.1.
REFERENCES None
ADVs B 3.7.18 B 3.7 PLANT SYSTEMS B 3.7.18 Atmospheric Dump Valves (ADVs)
BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.18-1 Revision 54 BACKGROUND The ADVs provide a safety grade method for cooling the unit to Shutdown Cooling (SDC) System entry conditions, should the preferred heat sink via the Turbine Bypass Valves to the condenser not be available, as discussed in the UFSAR, Section 10.3 (Reference 1). This is done in conjunction with the Auxiliary Feedwater System providing cooling water from the condensate storage tank (CST). The ADVs may also be used during a normal cooldown when steam pressure drops too low for maintenance of a vacuum in the condenser to permit use of the Turbine Bypass Valves.
Two ADV lines are provided, one per steam generator. Each ADV line consists of one ADV and an associated isolation valve. The ADVs are provided with upstream isolation valves to permit their being tested at power, if desired. The ADVs are equipped with manual hand wheels to open and close them.
Pneumatic controllers are used to operate the ADVs as the preferred method, but are not relied upon during an accident. A description of the ADVs is found in Reference 1. The ADVs are considered OPERABLE when the manual control is available for local manual operation.
APPLICABLE The design basis of the ADVs is established by the SAFETY ANALYSES capability to cool the unit to SDC System entry conditions.
The cooldown rate assumed in the accident analyses is obtainable by one or both steam generators. The design is adequate to cool the unit to SDC System entry conditions with only one ADV and one steam generator.
In the steam generator tube rupture accident analysis presented in the UFSAR, the ADVs are assumed to be used by the operator to cool down the unit to SDC System entry conditions because the accident is accompanied by a loss of offsite power. Prior to the operator action, the MSSVs are used to maintain steam generator pressure and temperature at the MSSV setpoint. The ADVs may be used for other accidents that are accompanied by a loss of offsite power. The limiting events are those that render one steam generator unavailable for RCS heat removal, with a coincident loss of offsite power. Typical initiating events falling into this ADVs B 3.7.18 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.18-2 Revision 54 category are a feedwater line break, and a SGTR event (limiting case).
The ADVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO Two ADV lines are required to be OPERABLE to ensure that at least one ADV is OPERABLE to conduct a unit cooldown following an event in which one steam generator becomes unavailable. A closed isolation valve does not render its ADV line inoperable since operator action time to open the isolation valve is supported in the accident analysis.
Failure to meet the LCO can result in the inability to cool the unit to SDC System entry conditions following an event in which the condenser is unavailable for use with the Turbine Bypass Valves. An ADV is considered OPERABLE when it is capable of providing relief of the main steam flow, and is capable of fully opening and closing when required.
APPLICABILITY In MODES 1, 2, and 3, and in MODE 4, when steam generators are being relied upon for heat removal, the ADVs are required to be OPERABLE. In MODES 5 and 6, a SGTR is not a credible event.
ACTIONS A.1 With one required ADV line inoperable, action must be taken to restore the OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> Completion Time takes into account the redundant capability afforded by the remaining OPERABLE ADV line, and a backup in the Turbine Bypass Valves and MSSVs.
B.1 With two required ADV lines inoperable, action must be taken to restore one of the ADV lines to OPERABLE status. As the isolation valve can be closed to isolate an ADV, some repairs may be possible with the unit at power. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is reasonable to repair inoperable ADV lines, based on the availability of the Turbine Bypass Valves and MSSVs, and the low probability of an event occurring during this period that requires the ADV lines.
ADVs B 3.7.18 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.7.18-3 Revision 55 C.1 and C.2 If the ADV lines cannot be restored to OPERABLE status
within the associated Completion Time, the unit must be
placed in a MODE in which the LCO does not apply. To
achieve this status, the unit must be placed in at least
MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4, without reliance upon
the steam generator for heat removal, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The
allowed Completion Times are reasonable, based on operating
experience, to reach the required unit conditions from full
power conditions in an orderly manner and without
challenging unit systems.
SURVEILLANCE SR 3.7.18.1 REQUIREMENTS To perform a cooldown of the RCS, the ADVs must be able to be opened through their full range. This SR ensures the
ADVs are tested through a full cycle at least once per fuel
cycle. This test is performed using the manual handwheel
assembly. Any use of an ADV using the manual handwheel
assembly may satisfy this requirement. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES 1. UFSAR, Section 10.3