ML16258A084
| ML16258A084 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 09/08/2016 |
| From: | Exelon Generation Co |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML16258A079 | List: |
| References | |
| Download: ML16258A084 (29) | |
Text
Boron Concentration B 3.9.1 B 3.9 REFUELING OPERATIONS
B 3.9.1 Boron Concentration
BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.1-1 Revision 11 BACKGROUND The limit on the boron concentrations of the Reactor Coolant System (RCS) and refueling pool during refueling, ensures
that the reactor remains subcritical during MODE 6.
Refueling boron concentration is the soluble boron concentration in the coolant in each of these volumes that have direct access to the reactor core during refueling.
The soluble boron concentration offsets the core reactivity and is measured by chemical analysis of a representative
sample of the coolant in each of the volumes. The refueling
boron concentration limit is specified in the Core Operating
Limits Report (COLR). Unit procedures ensure the specified
boron concentration in order to maintain an overall core reactivity of keff 0.95 during fuel handling, with control element assemblies and fuel assemblies assumed to be in the
most adverse configuration (least negative reactivity)
allowed by unit procedures. The negative worth of the CEAs
may be credited when determining the refueling boron
concentrations. Unit procedures maintain the number and position of credited CEAs during fuel handling operations.
Reference 1, Appendix 1C, Criterion 27, requires that two independent reactivity control systems of different design
principles be provided. One of these systems must be
capable of holding the reactor core subcritical under cold
conditions. The Chemical and Volume Control System is the
system capable of maintaining the reactor subcritical in
cold conditions by maintaining the boron concentration.
The reactor is brought to shutdown conditions before beginning operations to open the reactor vessel for
refueling. After the RCS is cooled and depressurized and
the vessel head is unbolted, the head is slowly removed to form the refueling pool. The refueling pool is then flooded with borated water from the refueling water tank into the
open reactor vessel by gravity feeding or by the use of the
Shutdown Cooling (SDC) System pumps.
Boron Concentration B 3.9.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.1-2 Revision 13 The pumping action of the SDC System in the RCS and the natural circulation due to thermal driving heads in the
reactor vessel and the refueling pool mix the added
concentrated boric acid with the water in the RCS. The SDC
System is in operation during refueling [see Limiting
Condition of Operation (LCO) 3.9.4 and LCO 3.9.5], to provide forced circulation in the RCS and assist in maintaining the boron concentrations in the RCS and the
refueling pool above the COLR limit.
The COLR includes a MODE 6 temperature limitation of 140°F. This restriction ensures assumptions made for calculating boron concentration and assumptions made in the boron dilution analysis for MODE 6 are preserved.
APPLICABLE During refueling operations, the reactivity condition of the SAFETY ANALYSES core is consistent with the initial conditions assumed for the boron dilution accident in the accident analysis and is
conservative for MODE 6.
The required boron concentration and the unit refueling procedures that demonstrate the correct fuel loading plan (including full core mapping) ensure the keff of the core will remain 0.95 during the refueling operation. Hence, at least a 5% k/k margin of safety is established during refueling. The boron concentration limit specified in the COLR includes an uncertainty allowance.
During refueling, the water volume in the spent fuel pool, the transfer tube, the refueling pool, and the reactor vessel form a single mass. As a result, the soluble boron concentration is relatively the same in each of these
volumes.
The limiting boron dilution accident analyzed occurs in MODE 5 (Reference 1, Chapter 14). A detailed discussion of
this event is provided in B 3.1.1.
The RCS boron concentration satisfies 10 CFR 50.36(c)(2)(ii), Criterion 2.
Boron Concentration B 3.9.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.1-3 Revision 29 LCO The LCO requires that a minimum boron concentration be maintained in the RCS and refueling pool while in MODE 6.
The boron concentration limit specified in the COLR ensures a core keff of 0.95 is maintained during fuel handling operations. Violation of the LCO could lead to an inadvertent criticality during MODE 6.
APPLICABILITY This LCO is applicable in MODE 6 to ensure that the fuel in the reactor vessel will remain subcritical. The required boron concentration ensures a keff 0.95. Above MODE 6, LCO 3.1.1 ensures that an adequate amount of negative
reactivity is available to shut down the reactor and to maintain it subcritical.
ACTIONS A.1 Continuation of positive reactivity additions (including actions to reduce boron concentration) is contingent upon
maintaining the unit in compliance with the LCO. If the
boron concentration of any coolant volume in the RCS or the
refueling pool is less than its limit, all operations
involving positive reactivity additions must be suspended immediately.
Suspension of positive reactivity additions shall not preclude moving a component to a safe position.
A.2 In addition to immediately suspending positive reactivity additions, boration to restore the concentration must be
initiated immediately.
In determining the required combination of boration flow rate and concentration, there is no unique design basis
event that must be satisfied. The only requirement is to
restore the boron concentration to its required value as
soon as possible. In order to raise the boron concentration
as soon as possible, the operator should begin boration with
the best source available for unit conditions.
Once boration is initiated, it must be continued until the boron concentration is restored. The restoration time Boron Concentration B 3.9.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.1-4 Revision 55 depends on the amount of boron that must be injected to reach the required concentration.
SURVEILLANCE SR 3.9.1.1 REQUIREMENTS This Surveillance Requirement (SR) ensures the coolant boron concentration in the RCS and the refueling pool is within
the COLR limits. The coolant boron concentration in each
volume is determined periodically by chemical analysis.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES 1. Updated Final Safety Analysis Report (UFSAR)
Nuclear Instrumentation B 3.9.2 B 3.9 REFUELING OPERATIONS
B 3.9.2 Nuclear Instrumentation
BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.2-1 Revision 32 BACKGROUND The source range monitors (SRMs) are used during refueling operations to monitor the core reactivity condition. The
installed SRMs are part of the wide range nuclear
instrumentation which are part of the Nuclear Instrumentation System. These detectors are located external to the reactor vessel and detect neutrons leaking
from the core. The use of portable detectors is permitted, provided the LCO requirements are met.
The installed SRMs are fission chambers. The detectors monitor the neutron flux in counts per second. The
instrument range (shutdown monitors) covers five decades of neutron flux (1E+5 cps) with at least a
+/- 5% instrument accuracy. The detectors also provide continuous visual
indication in the Control Room and an audible indication in
the Control Room and Containment to alert operators to a
possible dilution accident. The Nuclear Instrumentation
System is designed in accordance with the criteria presented
in Reference 1, Appendix 1C.
If used, portable detectors should be functionally equivalent to the Nuclear Instrumentation System SRMs.
APPLICABLE Two OPERABLE SRMs are required to provide a signal to alert SAFETY ANALYSES the operator to unexpected changes in core reactivity such as by a boron dilution accident or an improperly loaded fuel
assembly. The safety analysis of the uncontrolled boron
dilution accident is described in Reference 1, Chapter 14.
The analysis of the uncontrolled boron dilution accident shows that normally available SHUTDOWN MARGIN would be
reduced, but there is sufficient time for the operator to
take corrective actions.
The SRMs satisfy 10 CFR 50.36(c)(2)(ii), Criterion 3.
LCO This LCO requires two SRMs OPERABLE to ensure that redundant monitoring capability is available to detect changes in core reactivity.
Nuclear Instrumentation B 3.9.2 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.2-2 Revision 29 APPLICABILITY In MODE 6, the SRMs must be OPERABLE to determine changes in core reactivity. There is no other direct means available
to check core reactivity levels.
In MODEs 2, 3, 4, and 5, the installed source range detectors and circuitry are required to be OPERABLE by
ACTIONS A.1 and A.2 With only one SRM OPERABLE, redundancy has been lost. Since these instruments are the only direct means of monitoring
core reactivity conditions, positive reactivity additions and introduction of water into the RCS with a boron
concentration less than that required to meet the minimum
boron concentration of LCO 3.9.1 must be suspended
immediately. Suspending positive reactivity additions that
could result in failure to meet the minimum boron
concentration limit is required to assure continued safe
operation. Introduction of coolant inventory must be from
sources that have a boron concentration greater than that
required in the RCS for the minimum refueling boron
concentration. This may result in an overall reduction in
RCS boron concentration, but provides acceptable margin to
maintaining subcritical operation. Performance of Required
Action A.1 shall not preclude completion of movement of a
component to a safe position.
B.1 With no SRM OPERABLE, action to restore a monitor to OPERABLE status shall be initiated immediately. Once initiated, action shall be continued until an SRM is
restored to OPERABLE status.
B.2 With no SRM OPERABLE, there is no direct means of detecting changes in core reactivity. However, since positive reactivity additions are not to be made, the core reactivity
condition is stabilized until the SRMs are OPERABLE. This
stabilized condition is determined by performing SR 3.9.1.1
to verify that the required boron concentration exists.
Nuclear Instrumentation B 3.9.2 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.2-3 Revision 55 The Completion Time of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient to obtain and analyze a reactor coolant sample for boron
concentration and ensures that unplanned changes in boron
concentration would be identified. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is
reasonable, considering the low probability of a change in core reactivity during this period.
SURVEILLANCE SR 3.9.2.1 REQUIREMENTS Surveillance Requirement 3.9.2.1 is the performance of a CHANNEL CHECK, which is a comparison of the parameter
indicated on one channel to a similar parameter on other
channels. It is based on the assumption that the two
indication channels should be consistent with core
conditions. Changes in fuel loading and core geometry can
result in significant differences between source range
channels, but each channel should be consistent with its
local conditions.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.9.2.2 Surveillance Requirement 3.9.2.2 is the performance of a CHANNEL CALIBRATION. This SR is modified by a Note stating that neutron detectors are excluded from the CHANNEL
CALIBRATION. This is because generating a meaningful test
signal is difficult; the detectors are of simple
construction, and any failures in the detectors will be
apparent as a change in channel output. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES 1. UFSAR
Containment Penetrations B 3.9.3 B 3.9 REFUELING OPERATIONS
B 3.9.3 Containment Penetrations
BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.3-1 Revision 41 BACKGROUND During movement of irradiated fuel assemblies within the Containment Structure, a release of fission product
radioactivity within the Containment Structure will be
restricted from escaping to the environment when the LCO requirements are met. In MODEs 1, 2, 3, and 4, this is accomplished by maintaining Containment OPERABLE, as
described in LCO 3.6.1. In MODE 6, the potential for
containment pressurization as a result of an accident is not
likely; therefore, requirements to isolate the containment
atmosphere from the outside atmosphere can be less
stringent. The LCO requirements are referred to as "containment closure" rather than "containment OPERABILITY."
Containment closure means that all potential filtered or
unfiltered escape paths are closed or capable of being
closed. Since there is no design basis accident potential
for containment pressurization, the Appendix J leakage
criteria and tests are not required.
The Containment Structure serves to contain fission product radioactivity that may be released from the reactor core
following an accident, such that offsite radiation exposures
are maintained well within the requirements of 10 CFR 50.67.
Additionally, the Containment Structure provides radiation
shielding from the fission products that may be present in
the containment atmosphere following accident conditions.
The containment equipment hatch opening provides a means for moving large equipment and components into and out of the
Containment Structure. During movement of irradiated fuel
assemblies within Containment, the equipment hatch must be
held in place by at least four bolts or the containment
outage door must be capable of being closed. Good engineering practice dictates that the bolts required by this LCO be approximately equally spaced.
The containment air locks, which are part of the containment pressure boundary, provide a means for personnel access
during MODEs 1, 2, 3, and 4 operation in accordance with
LCO 3.6.2. Each air lock has a door at both ends. The Containment Penetrations B 3.9.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.3-2 Revision 41 doors are normally interlocked to prevent simultaneous opening when Containment OPERABILITY is required.
In other situations, the potential for containment pressurization as a result of an accident is not present, therefore, less stringent requirements are needed to isolate
the containment atmosphere from the outside atmosphere.
Both containment personnel air lock doors and the
containment outage door may be open during the movement of
irradiated fuel assemblies in containment; provided one air
lock door and the containment outage door are OPERABLE, the plant is in MODE 6 with at least 23 ft of water above the fuel, and a designated individual is continuously available
to close each door. The designated individuals must be
stationed at the Auxiliary Building side of the outer air
lock door and near the containment equipment hatch.
OPERABILITY of a containment personnel air lock door
requires that the door is capable of being closed, that the
door is unblocked, and no cables or hoses are run through
the doorway. OPERABILITY of the containment outage door
requires that the door is capable of being closed, that the
door is unblocked (i.e., capable of being closed within 30
minutes), and no cables and hoses are run through the
doorway. Containment outage door grating or truck ramps may
be installed if the grating or truck ramps can be removed
with the use of a forklift and the door closed within 30
minutes. Penetration flow paths with direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative control during movement of irradiated fuel assemblies in Containment provided that the appropriate personnel are aware of the open status and are designated and readily available to isolate the flow path in the event of a fuel handling accident. During movement of irradiated fuel assemblies in containment, the requirement
for at least 23 ft of water above the fuel, ensures that
there is sufficient time to close the personnel air lock, the containment outage door, and penetration flow paths following a loss of SDC before boiling occurs and minimizes
activity release after a fuel handling accident. The personnel air lock door, the containment outage door, and penetration flow paths may be operated independently of each other (i.e., they do not have to be open or shut at the same time).
Containment Penetrations B 3.9.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.3-3 Revision 58 The requirements on containment penetration closure, ensure that a release of fission product radioactivity within the
Containment Structure will be restricted to within
regulatory limits.
The Containment Purge Valve Isolation System, for the purposes of compliance with LCO 3.9.3, item d.2, includes a
48 inch purge penetration and a 48 inch exhaust penetration.
For the purposes of compliance with LCO 3.9.3, the
containment vent isolation valves are not considered part of
the Containment Purge Valve Isolation System since they may not be capable of being closed automatically. The containment vent, includes a four inch purge penetration and
a four inch exhaust penetration. During MODEs 1, 2, 3, and
4, the normal purge and exhaust penetrations are isolated (via a blind flange, if installed or by the purge valves).
The containment vent valves can be opened intermittently, but are closed automatically by the Engineered Safety
Features Actuation System. Neither of the subsystems is
subject to a Specification in MODE 5.
In MODE 6, large air exchanges are desired to conduct refueling operations. The normal 48 inch purge system is
used for this purpose and all valves are closed by the
Engineered Safety Features Actuation System in accordance
with LCO 3.3.7.
The other containment penetrations that provide direct access from containment atmosphere to outside atmosphere
through a filtered or unfiltered pathway must be isolated on
at least one side. Isolation may be achieved by an OPERABLE
automatic isolation valve, or by a manual isolation valve, blind flange, or equivalent. Equivalent isolation methods
must be approved in accordance with appropriate American
Society of Mechanical Engineers / American National
Standards Institute Codes, and may include use of a material
that can provide a temporary ventilation barrier for the other containment penetrations during fuel movements.
APPLICABLE During movement of irradiated fuel assemblies within the SAFETY ANALYSES Containment Structure, the most severe radiological consequences result from a fuel handling accident. The fuel
handling accident is a postulated event that involves damage
to irradiated fuel (Reference 1). The fuel handling Containment Penetrations B 3.9.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.3-4 Revision 58 accident, described in Reference 1, includes dropping a single irradiated fuel assembly which would then rotate to a
horizontal position, strike a protruding structure, and
rupture the fuel pins. The requirements of LCO 3.9.6, and
the minimum decay time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to movement of irradiated fuel assemblies, ensure that the release of
fission product radioactivity, subsequent to a fuel handling
accident, results in doses that are within the acceptance
limits given in Reference 1.
Containment penetrations satisfy 10 CFR 50.36(c)(2)(ii), Criterion 3.
LCO This LCO limits the consequences of a fuel handling accident in Containment Structure, by limiting the potential escape
paths for fission product radioactivity released within
Containment. The LCO requires any penetration providing
direct access from the Containment Structure atmosphere to
the outside atmosphere through a filtered or unfiltered
pathway (including the containment vent isolation valves) to
be closed, except for the OPERABLE containment purge and
exhaust penetrations, the containment personnel air locks
and the containment outage door. For the OPERABLE
containment purge and exhaust penetrations, this LCO ensures
that these penetrations are isolable by the Containment
Purge Valve Isolation System. The OPERABILITY requirements
for this LCO ensure that the automatic purge and exhaust
valve closure times specified in the UFSAR can be achieved, and therefore meet the assumptions used in the safety
analysis to ensure releases through the valves are
terminated, such that the radiological doses are within the
acceptance limit.
Both containment personnel air lock doors and the containment outage door may be open under administrative controls during movement of irradiated fuel assemblies in
Containment provided that one OPERABLE personnel air lock
door and an OPERABLE containment outage door are capable of
being closed in the event of a fuel handling accident. The
administrative controls consist of designated individuals
available immediately outside the personnel air lock and
near the containment outage door to close the OPERABLE
doors. Should a fuel handling accident occur inside the
Containment Structure, one personnel air lock door and the Containment Penetrations B 3.9.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.3-5 Revision 58 containment outage door will be closed following an evacuation of the Containment Structure.
The LCO is modified by a Note which allows the emergency air lock temporary closure device to replace an emergency air
lock door. The temporary closure device provides an
adequate barrier to shield the environment from the
containment atmosphere in case of a design basis event that
does not create a pressure increase inside Containment.
The LCO is modified by a Note allowing penetration flow paths with direct access from the containment atmosphere to the outside atmosphere to be unisolated under administrative
controls. Administrative controls ensure that
- 1) appropriate personnel are aware of the open status of the
penetration flow path during movement of irradiated fuel
assemblies within Containment, and 2) specified individuals
are designated and readily available to isolate the flow path in the event of a fuel handling accident.
APPLICABILITY The containment penetration requirements are applicable during movement of irradiated fuel assemblies within the
Containment Structure because this is when there is a
potential for a fuel handling accident. In MODEs 1, 2, 3, and 4, containment penetration requirements are addressed by
LCO 3.6.1. In MODEs 5 and 6, when movement of irradiated
fuel assemblies within the Containment Structure are not
being conducted, the potential for a fuel handling accident
does not exist. Therefore, under these conditions no requirements are placed on containment penetration status.
ACTIONS A.1 With the containment equipment hatch, air locks, or any containment penetration that provides direct access from the
containment atmosphere to the outside atmosphere through a
filtered or unfiltered pathway not in the required status, (including the Containment Purge and Exhaust Isolation
System not capable of automatic actuation when the purge and
exhaust valves are open) the unit must be placed in a
condition in which the isolation function is not needed.
This is accomplished by immediately suspending movement of
irradiated fuel assemblies within the Containment Structure.
Containment Penetrations B 3.9.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.3-6 Revision 58 Performance of these actions shall not preclude completion of movement of a component to a safe position.
SURVEILLANCE SR 3.9.3.1 REQUIREMENTS This SR demonstrates that each of the containment penetrations required to be in its closed position, is in
that position. The surveillance test on the open purge and
exhaust valves will demonstrate that the valves are not
blocked from closing. Also, the surveillance test will
demonstrate that each purge and exhaust valve operator has motive power, which will ensure each valve is capable of being closed by an OPERABLE automatic Containment Purge
Valve Isolation System.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.9.3.2 This SR demonstrates that each containment purge and exhaust valve actuates to its isolation position on manual
initiation or on an actual or simulated high radiation
signal. The Surveillance Frequency is controlled under the
Surveillance Frequency Control Program. These surveillance
tests performed during MODE 6 will ensure that the valves
are capable of closing after a postulated fuel handling
accident to limit a release of fission product radioactivity from the Containment Structure.
REFERENCES 1. UFSAR, Section 14.18, "Fuel Handling Incident" SDC and Coolant Circulation-High Water Level B 3.9.4 B 3.9 REFUELING OPERATIONS
B 3.9.4 Shutdown Cooling (SDC) and Coolant Circulation-High Water Level
BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.4-1 Revision 2 BACKGROUND The purposes of the SDC System in MODE 6 are to remove decay heat and other residual heat from the RCS, to provide mixing of borated coolant, to provide sufficient coolant
circulation to minimize the effects of a boron dilution accident, and to prevent boron stratification (Reference 1).
Heat is removed from the RCS by circulating reactor coolant
through the SDC heat exchanger(s), where the heat is
transferred to the Component Cooling Water System via the
SDC heat exchanger(s). The coolant is then returned to the
RCS via the RCS cold leg(s). Operation of the SDC System
for normal cooldown or decay heat removal is manually
accomplished from the Control Room. The heat removal rate is adjusted by controlling the flow of reactor coolant
through the SDC heat exchanger(s) and bypassing the heat
exchanger(s). Mixing of the reactor coolant is maintained
by this continuous circulation of reactor coolant through the SDC System.
APPLICABLE If the reactor coolant temperature is not maintained below SAFETY ANALYSES 200°F, boiling of the reactor coolant could result. This could lead to inadequate cooling of the reactor fuel due to
a resulting loss of coolant in the reactor vessel.
Additionally, boiling of the reactor coolant could lead to a
reduction in boron concentration in the coolant, due to the boron plating out on components near the areas of the
boiling activity, and due to the possible addition of water to the reactor vessel with a lower boron concentration than
is required to keep the reactor subcritical. The loss of reactor coolant and the reduction of boron concentration in the reactor coolant would eventually challenge the integrity
of the fuel cladding, which is a fission product barrier.
One loop of the SDC System is required to be operational in MODE 6, with the water level 23 ft above the top of the irradiated fuel assemblies seated in the reactor vessel, to
prevent this challenge. The LCO does permit de-energizing
of the SDC pump for short durations under the condition that
the boron concentration is not diluted. This conditional
de-energizing of the SDC pump does not result in a challenge
to the fission product barrier.
SDC and Coolant Circulation-High Water Level B 3.9.4 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.4-2 Revision 56 Shutdown cooling and Coolant Circulation-High Water Level satisfies 10 CFR 50.36(c)(2)(ii), Criterion 2.
LCO Only one SDC loop is required for decay heat removal in MODE 6, with water level 23 ft above the top of the irradiated fuel assemblies seated in the reactor vessel.
Only one SDC loop is required because the volume of water
above the irradiated fuel assemblies seated in the reactor
vessel provides backup decay heat removal capability. At
least one SDC loop must be OPERABLE and in operation to
provide: a. Removal of decay heat;
- b. Mixing of borated coolant to minimize the possibility of a criticality; and c. Indication of reactor coolant temperature.
An OPERABLE SDC loop includes an SDC pump, a heat exchanger, valves, piping, instruments, and controls to ensure an
OPERABLE flow path, and to determine the low end
temperature. An OPERABLE SDC loop is supported by a
functional saltwater and component cooling water subsystem.
Note that one functional saltwater and component cooling
water subsystem may support both OPERABLE SDC loops, if its
heat removal capacity is sufficient. The flow path starts
in one of the RCS hot legs and is returned to the RCS cold
legs. Management of gas voids is important to SDC System OPERABILITY.
The LCO is modified by a Note that allows the required operating SDC loop not to be in operation for up to one hour in each eight hour period, provided no operations are permitted that would cause the introduction of water into
the RCS with a boron concentration less than that required
to meet the minimum boron concentration of LCO 3.9.1. Boron
concentration reduction with water at boron concentrations
less than that required to meet the minimum boron
concentration of LCO 3.9.1 is prohibited because uniform
concentration distribution cannot be ensured without forced
circulation. This permits operations such as core mapping
or alterations in the vicinity of the reactor vessel hot leg
nozzles, and RCS to SDC isolation valve testing. During SDC and Coolant Circulation-High Water Level B 3.9.4 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.4-3 Revision 56 this one hour period, decay heat is removed by natural convection to the large mass of water in the refueling pool.
A second Note also allows both SDC loops not to be in operation during the time required for local leak rate
testing of Containment Penetration Number 41 pursuant to the requirements of SR 3.6.1.1 or to permit maintenance on valves located in the common SDC suction line. In addition
to the requirement in Note 1 regarding control of boron
concentration, movement of fuel assemblies is suspended
within Containment and all containment penetrations must be
in the status described in LCO 3.9.3. This allowance is
necessary to perform required maintenance and testing.
APPLICABILITY One SDC loop must be in operation in MODE 6, with the water level 23 ft above the top of the irradiated fuel assemblies seated in the reactor vessel, to provide decay
heat removal. The 23 ft level was selected because it
corresponds to the 23 ft requirement established for fuel
movement in LCO 3.9.6.
Requirements for the SDC System in other MODEs are covered by LCOs in Section 3.4 and Section 3.5. Shutdown cooling
loop requirements in MODE 6, with the water level < 23 ft
above the top of the irradiated fuel assemblies seated in the reactor vessel, are located in LCO 3.9.5.
ACTIONS Shutdown cooling loop requirements are met by having one SDC loop OPERABLE and in operation, except as permitted in the
Note to the LCO.
A.1 If one required SDC loop is inoperable or not in operation, action shall be immediately initiated and continued until
the SDC loop is restored to OPERABLE status and to
operation. An immediate Completion Time is necessary for an
operator to initiate corrective actions.
A.2 If SDC loop requirements are not met, there will be no forced circulation to provide mixing to establish uniform
boron concentrations. Suspending positive reactivity SDC and Coolant Circulation-High Water Level B 3.9.4 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.4-4 Revision 56 additions that could result in failure to meet the minimum boron concentration limit is required to assure continued
safe operation. Introduction of coolant inventory must be
from sources that have a boron concentration greater than
that required in the RCS for the minimum refueling boron
concentration. This may result in an overall reduction in RCS boron concentration, but provides an acceptable margin to maintaining subcritical operation. In addition, to
ensure compliance with the action is maintained, the
charging pumps shall be de-energized and charging flow paths
closed as part of Required Action A.2.
A.3 If SDC loop requirements are not met, actions shall be taken immediately to suspend loading irradiated fuel assemblies in
the core. With no forced circulation cooling, decay heat
removal from the core occurs by natural convection to the
heat sink provided by the water above the core. A minimum
refueling water level of 23 ft above the irradiated fuel
assemblies seated in the reactor vessel provides an adequate
available heat sink. Suspending any operation that would
increase the decay heat load, such as loading a fuel
assembly, is a prudent action under this condition.
A.4.1, A.4.2, A.5, A.6.1, and A.6.2 If no SDC is in operation, the following actions must be
taken: a. the equipment hatch must be closed and secured with a minimum of four bolts or the containment outage door
must be closed; b. one door in each air lock must be closed; and
- c. each penetration providing direct access from the containment atmosphere to the outside atmosphere must
be either closed by a manual or automatic isolation valve, blind flange, or equivalent, or verified to be capable of being closed by an OPERABLE containment
purge valve isolation system. This requirement is to
be applied to each penetration separately. The
decision to apply A.6.1 or A.6.2 is made for each
penetration, depending on whether or not that SDC and Coolant Circulation-High Water Level B 3.9.4 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.4-5 Revision 56 penetration has an OPERABLE containment purge valve isolation system.
With SDC loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the
containment atmosphere. Performing the actions described above ensure that all containment penetrations are either closed or can be closed so that the dose limits are not
exceeded.
The Completion Time of four hours allows fixing of most SDC problems and is reasonable, based on the low probability of
the coolant boiling in that time.
The emergency air lock temporary closure device cannot be credited for containment closure for a loss of shutdown
cooling event. At least one door in the emergency air lock must be closed to satisfy this action statement.
SURVEILLANCE SR 3.9.4.1 REQUIREMENTS This SR demonstrates that the SDC loop is in operation and circulating reactor coolant. The flow rate is determined by
the flow rate necessary to provide sufficient decay heat
removal capability, and to prevent thermal and boron
stratification in the core. The Surveillance Frequency is
controlled under the Surveillance Frequency Control Program.
SR 3.9.4.2 SDC System piping and components have the potential to develop voids and pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessary for proper operation of the required SDC loop(s) and may also prevent water hammer, pump cavitation, and pumping of noncondensible gas into the reactor vessel.
Selection of SDC System locations susceptible to gas accumulation is based on a review of system design information, including piping and instrumentation drawings, isometric drawings, plan and elevation drawings, and calculations. The design review is supplemented by system walk downs to validate the system high points and to confirm the location and orientation of important components that SDC and Coolant Circulation-High Water Level B 3.9.4 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.4-6 Revision 58 can become sources of gas or could otherwise cause gas to be trapped or difficult to remove during system maintenance or
restoration. Susceptible locations depend on plant and
system configuration, such as stand-by versus operating
conditions.
The SDC System is OPERABLE when it is sufficiently filled with water. Acceptance criteria are established for the
volume of accumulated gas at susceptible locations. If
accumulated gas is discovered that exceeds the acceptance
criteria for the susceptible location (or the volume of
accumulated gas at one or more susceptible locations exceeds
an acceptance criteria for gas volume at the suction or
discharge of a pump), the Surveillance is not met. If it is
determined by subsequent evaluation that the SDC System is
not rendered inoperable by the accumulated gas (i.e., the
system is sufficiently filled with water), the Surveillance
may be declared met. Accumulated gas should be eliminated
or brought within the acceptance criteria limits.
SDC System locations susceptible to gas accumulation are
monitored and, if gas is found, the gas volume is compared
to the acceptance criteria for the location. Susceptible
locations in the same system flow path which are subject to
the same gas intrusion mechanisms may be verified by
monitoring a representative sub-set of susceptible
locations. Monitoring may not be practical for locations
that are inaccessible due to radiological or environmental
conditions, the plant configuration, or personnel safety.
For these locations alternative methods (e.g., operating
parameters, remote monitoring) may be used to monitor the
susceptible location. Monitoring is not required for
susceptible locations where the maximum potential
accumulated gas void volume has been evaluated and determined to not challenge system OPERABILITY. The accuracy of the method used for monitoring the susceptible
locations and trending of the results should be sufficient
to assure system OPERABILITY during the Surveillance
interval.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SDC and Coolant Circulation-High Water Level B 3.9.4 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.4-7 Revision 56 REFERENCES 1. UFSAR, Section 9.2, "Shutdown Cooling System" SDC and Coolant Circulation-Low Water Level B 3.9.5 B 3.9 REFUELING OPERATIONS
B 3.9.5 Shutdown Cooling (SDC) and Coolant Circulation-Low Water Level
BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.5-1 Revision 2 BACKGROUND The purposes of the SDC System in MODE 6 are to remove decay heat and other residual heat from the RCS, to provide mixing of borated coolant, to provide sufficient coolant
circulation to minimize the effects of a boron dilution accident, and to prevent boron stratification (Reference 1).
Heat is removed from the RCS by circulating reactor coolant
through the SDC heat exchanger(s), where the heat is
transferred to the Component Cooling Water System via the
SDC heat exchanger(s). The coolant is then returned to the
RCS via the RCS cold leg(s). Operation of the SDC System
for normal cooldown or decay heat removal is manually
accomplished from the Control Room. The heat removal rate is adjusted by controlling the flow of reactor coolant
through the SDC heat exchanger(s) and bypassing the heat
exchanger(s). Mixing of the reactor coolant is maintained
by this continuous circulation of reactor coolant through the SDC System.
APPLICABLE If the reactor coolant temperature is not maintained below SAFETY ANALYSES 200°F, boiling of the reactor coolant could result. This could lead to inadequate cooling of the reactor fuel due to
the resulting loss of coolant in the reactor vessel.
Additionally, boiling of the reactor coolant could lead to a
reduction in boron concentration in the coolant due to the
boron plating out on components near the areas of the
boiling activity, and due to the possible addition of water to the reactor vessel with a lower boron concentration than
is required to keep the reactor subcritical. The loss of reactor coolant and the reduction of boron concentration in the reactor coolant would eventually challenge the integrity
of the fuel cladding, which is a fission product barrier.
Two loops of the SDC System are required to be OPERABLE, and
one loop is required to be in operation in MODE 6, with the
water level < 23 ft above the top of the irradiated fuel
assemblies seated in the reactor vessel, to prevent this
challenge.
Shutdown cooling and Coolant Circulation-Low Water Level satisfies 10 CFR 50.36(c)(2)(ii), Criterion 2.
SDC and Coolant Circulation-Low Water Level B 3.9.5 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.5-2 Revision 56 LCO In MODE 6, with the water level < 23 ft above the top of the irradiated fuel assemblies seated in the reactor vessel, both SDC loops must be OPERABLE. Additionally, one loop of
the SDC System must be in operation in order to provide: a. Removal of decay heat;
- b. Mixing of borated coolant to minimize the possibility of a criticality; and c. Indication of reactor coolant temperature.
An OPERABLE SDC loop consists of an SDC pump, a heat exchanger, valves, piping, instruments, and controls to
ensure an OPERABLE flow path and to determine the low end
temperature. An OPERABLE SDC loop is supported by a
functional saltwater and component cooling water subsystem.
Note that one functional saltwater and component cooling
water subsystem may support both OPERABLE SDC loops, if its
heat removal capacity is sufficient. The flow path starts
in one of the RCS hot legs and is returned to the RCS cold
legs. Both SDC pumps may be aligned to the Refueling Water
Tank to support filling the refueling pool or for
performance of required testing. Management of gas voids is important to SDC System OPERABILITY.
The LCO is modified by a note that allows one required SDC loop to be replaced by one spent fuel pool cooling loop when
it is lined up to provide cooling flow to the irradiated
fuel assemblies in the reactor core, and the heat generation
rate of the core is below the heat removal capacity of the
spent fuel cooling loop.
This LCO is modified by a Note that allows one SDC loop to be inoperable for a period of two hours provided the other
loop is OPERABLE and in operation. Prior to declaring the
loop inoperable, consideration should be given to the existing plant configuration. This consideration should include that the core time to boil is short, there is no
draining operation to further reduce RCS water level and
that the capability exists to inject borated water into the
reactor vessel. This permits surveillance tests to be
performed on the inoperable loop during a time when these
tests are safe and possible.
SDC and Coolant Circulation-Low Water Level B 3.9.5 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.5-3 Revision 41 This LCO is modified by a Note that permits the SDC pumps to be deenergized for 15 minutes when switching from one train to another. The circumstances for stopping both SDC
pumps are to be limited to situations when the pump outage
time is short and the core outlet temperature is maintained 10F below saturation temperature. The Note prohibits introduction of water with a boron concentration less than that required by LCO 3.9.1 or draining operations when SDC forced flow is stopped.
APPLICABILITY Two SDC loops are required to be OPERABLE, and one SDC loop must be in operation in MODE 6, with the water level < 23 ft
above the top of the irradiated fuel assemblies seated in
the reactor vessel, to provide decay heat removal.
Requirements for the SDC System in other MODEs are covered
by LCOs in Section 3.4. MODE 6 requirements, with a water level 23 ft above the irradiated fuel assemblies seated in the reactor vessel, are covered in LCO 3.9.4.
ACTIONS A.1 and A.2 If one SDC loop is inoperable, action shall be immediately initiated and continued until the SDC loop is restored to
OPERABLE status and is in operation, or until the water level is 23 ft above the irradiated fuel assemblies seated in the reactor vessel. When the water level is established at 23 ft above the irradiated fuel assemblies seated in the reactor vessel, the Applicability will change
to that of LCO 3.9.4, and only one SDC loop is required to
be OPERABLE and in operation. An immediate Completion Time
is necessary for an operator to initiate corrective actions.
B.1 If no SDC loop is in operation or no SDC loops are OPERABLE, there will be no forced circulation to provide mixing to
establish uniform boron concentrations. Suspending positive
reactivity additions that could result in failure to meet
the minimum boron concentration limit is required to assure
continued safe operation. Introduction of coolant inventory
must be from sources that have a boron concentration greater
than that required in the RCS for the minimum refueling
boron concentration. This may result in an overall
reduction in RCS boron concentration, but provides an SDC and Coolant Circulation-Low Water Level B 3.9.5 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.5-4 Revision 41 acceptable margin to maintaining subcritical operation. In addition, to ensure compliance with the action is
maintained, the charging pumps shall be de-energized and
charging flow paths closed as part of Required Action B.1.
B.2 If no SDC loop is in operation or no SDC loops are OPERABLE, action shall be initiated immediately and continued without
interruption to restore one SDC loop to OPERABLE status and
operation. Since the unit is in Conditions A and B
concurrently, the restoration of two OPERABLE SDC loops and
one operating SDC loop should be accomplished expeditiously.
B.3.1, B.3.2, B.4, B.5.1, and B.5.2 If no SDC is in operation, the following actions must be taken: a. the equipment hatch must be closed and secured with a minimum of four bolts or the containment outage door
must be closed; b. one door in each air lock must be closed; and
- c. each penetration providing direct access from the containment atmosphere to the outside atmosphere must
be either closed by a manual or automatic isolation
valve, blind flange, or equivalent, or verified to be
capable of being closed by an OPERABLE containment
purge valve isolation system. This requirement is to
be applied to each penetration separately. The
decision to apply B.5.1 or B.5.2 is made for each
penetration, depending on whether or not that
penetration has an OPERABLE containment purge valve
isolation system.
With SDC loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere. Performing the actions described above ensure that all containment penetrations are either
closed or can be closed so that the dose limits are not
exceeded.
SDC and Coolant Circulation-Low Water Level B 3.9.5 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.5-5 Revision 56 The Completion Time of four hours allows fixing of most SDC problems and is reasonable, based on the low probability of
the coolant boiling in that time.
The emergency air lock temporary closure device cannot be credited for containment closure for a loss of shutdown cooling event. At least one door in the emergency air lock
must be closed to satisfy this action statement.
SURVEILLANCE SR 3.9.5.1 REQUIREMENTS This SR demonstrates that one SDC loop is operating and circulating reactor coolant. The flow rate is determined by
the flow rate necessary to provide sufficient decay heat
removal capability and to prevent thermal and boron
stratification in the core. This SR also demonstrates that
the other SDC loop is OPERABLE.
In addition, during operation of the SDC loop with the water level in the vicinity of the reactor vessel nozzles, the SDC
loop flow rate determination must also consider the SDC pump
suction requirements. The Surveillance Frequency is
controlled under the Surveillance Frequency Control Program.
SR 3.9.5.2 This SR demonstrates that the SDC loop is in operation and circulating reactor coolant. The flow rate is determined by
the flow rate necessary to provide sufficient decay heat
removal capability and to prevent thermal and boron
stratification in the core. The Surveillance Frequency is
controlled under the Surveillance Frequency Control Program.
SR 3.9.5.3 Verification that the required pump and valves are OPERABLE ensures that an additional SDC loop can be placed in operation, if needed, to maintain decay heat removal and
reactor coolant circulation. Verification is performed by
verifying proper breaker alignment and power available to
the required pump and valves. The Surveillance Frequency is
controlled under the Surveillance Frequency Control Program.
SDC and Coolant Circulation-Low Water Level B 3.9.5 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.5-6 Revision 56 SR 3.9.5.4 SDC System piping and components have the potential to develop voids and pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessary for proper operation of the required SDC loops and may also prevent water hammer, pump cavitation, and pumping of noncondensible gas into the reactor vessel.
Selection of SDC System locations susceptible to gas accumulation is based on a review of system design information, including piping and instrumentation drawings, isometric drawings, plan and elevation drawings, and calculations. The design review is supplemented by system walk downs to validate the system high points and to confirm the location and orientation of important components that can become sources of gas or could otherwise cause gas to be trapped or difficult to remove during system maintenance or restoration. Susceptible locations depend on plant and system configuration, such as stand-by versus operating conditions.
The SDC System is OPERABLE when it is sufficiently filled with water. Acceptance criteria are established for the volume of accumulated gas at susceptible locations. If accumulated gas is discovered that exceeds the acceptance criteria for the susceptible location (or the volume of accumulated gas at one or more susceptible locations exceeds an acceptance criteria for gas volume at the suction or discharge of a pump), the Surveillance is not met. If it is determined by subsequent evaluation that the SDC System is not rendered inoperable by the accumulated gas (i.e., the system is sufficiently filled with water), the Surveillance may be declared met. Accumulated gas should be eliminated or brought within the acceptance criteria limits.
SDC System locations susceptible to gas accumulation are monitored and, if gas is found, the gas volume is compared to the acceptance criteria for the location. Susceptible locations in the same system flow path which are subject to the same gas intrusion mechanisms may be verified by monitoring a representative sub-set of susceptible locations. Monitoring may not be practical for locations that are inaccessible due to radiological or environmental SDC and Coolant Circulation-Low Water Level B 3.9.5 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.5-7 Revision 58 conditions, the plant configuration, or personnel safety.
For these locations alternative methods (e.g., operating
parameters, remote monitoring) may be used to monitor the
susceptible location. Monitoring is not required for
susceptible locations where the maximum potential
accumulated gas void volume has been evaluated and determined to not challenge system OPERABILITY. The accuracy of the method used for monitoring the susceptible
locations and trending of the results should be sufficient
to assure system OPERABILITY during the Surveillance
interval.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES 1. UFSAR, Section 9.2, "Shutdown Cooling System" Refueling Pool Water Level B 3.9.6 B 3.9 REFUELING OPERATIONS
B 3.9.6 Refueling Pool Water Level
BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.6-1 Revision 41 BACKGROUND The movement of irradiated fuel assemblies within the Containment Structure requires a minimum water level of
23 ft above the top of the irradiated fuel assemblies seated
in the reactor vessel. During refueling this maintains sufficient water level in the fuel transfer canal, the refueling pool, and the spent fuel pool. Sufficient water
is necessary to retain iodine fission product activity in
the water in the event of a fuel handling accident (References 1 and 2). Sufficient iodine activity would be
retained to limit offsite doses from the accident to within the acceptance criteria given in Reference 2.
APPLICABLE During movement of irradiated fuel assemblies, the water SAFETY ANALYSES level in the refueling pool is an initial condition design parameter in the analysis of the fuel handling accident in
the Containment Structure, postulated by Reference 1. A
minimum water level of 23 ft allows a decontamination factor of 200 (Appendix B, Section 2 of Reference 1) to be used in the accident analysis for iodine. This relates to the
assumption that 99.5% of the total iodine released from the pellet to cladding gap of all the dropped fuel assembly
rods, is retained by the refueling pool water. The fuel
pellet to cladding gap is assumed to contain 16% of I-131 and 10% of all other iodines (Reference 2).
The fuel handling accident analysis inside the Containment Structure is described in Reference 2. With a minimum water
level of 23 ft and a minimum decay time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to fuel handling, the analysis and test programs demonstrate
that the iodine release due to a postulated fuel handling
accident is adequately captured by the water and offsite
doses are maintained within allowable limits (Reference 2).
Refueling pool water level satisfies 10 CFR 50.36(c)(2)(ii), Criterion 2.
LCO A minimum refueling pool water level of 23 ft above the irradiated fuel assemblies seated in the reactor vessel is
required to ensure that the radiological consequences of a
postulated fuel handling accident inside the Containment Refueling Pool Water Level B 3.9.6 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.9.6-2 Revision 55 Structure are within the acceptable limits given in Reference 2.
APPLICABILITY LCO 3.9.6 is applicable when moving irradiated fuel assemblies in the Containment Structure. The LCO minimizes
the possibility of a fuel handling accident in the
Containment Structure that is beyond the assumptions of the
safety analysis. If irradiated fuel is not present in the
Containment Structure, there can be no significant
radioactivity release as a result of a postulated fuel
handling accident. Requirements for fuel handling accidents in the spent fuel pool are covered by LCO 3.7.13.
ACTIONS A.1 With a water level of < 23 ft above the top of the irradiated fuel assemblies seated in the reactor vessel, all
operations involving movement of irradiated fuel assemblies, shall be suspended immediately to ensure that a fuel
handling accident cannot occur.
The suspension of fuel movement shall not stop the movement of a component to a safe position.
SURVEILLANCE SR 3.9.6.1 REQUIREMENTS Verification of a minimum water level of 23 ft above the top of the irradiated fuel assemblies seated in the reactor
vessel ensures that the design basis for the postulated fuel
handling accident analysis during refueling operations is
met. Water at the required level above the top of the
irradiated fuel assemblies seated in the reactor vessel
limits the consequences of damaged fuel rods, that are
postulated to result from a fuel handling accident inside
the Containment Structure (Reference 2).
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES 1. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear
Power Reactors, July 2000 2. UFSAR, Section 14.18, "Fuel Handling Incident"