ML14267A231

B 3.6.1-1, Containment Systems Though B 3.6.8-4, Irs

ML14267A231
Person / Time
Site:	Calvert Cliffs
Issue date:	09/19/2014
From:	Calvert Cliffs, Exelon Generation Co
To:	Office of Nuclear Reactor Regulation
Shared Package
ML14267A237	List: ML14267A221 ML14267A223 ML14267A224 ML14267A225 ML14267A226 ML14267A227 ML14267A228 ML14267A229 ML14267A230 ML14267A231 ML14267A232 ML14267A233 ML14267A234
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Download: ML14267A231 (43)

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Containment B 3.6.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.1 Containment

BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.1-1 Revision 2 BACKGROUND The C ontainment Structure consists of the concrete building , its steel liner, and the penetrations through this structure. The structure is designed to contain radioactive material that may

be released from the reactor core following a Design Basis Accident (DBA). Additionally, this structure provides shielding from the fission products that may be present in the

containment atmosphere following accident conditions.

The C ontainment Structure is a reinforced concrete structure with a cylindrical wall, a flat foundation mat, and a shallow

dome roof. The C ontainment Structure has ungrouted tendons, therefore, the cylinder wall is prestressed with a

post-tensioning system in the vertical and horizontal directions, and the dome roof is prestressed utilizing a

three-way post-tensioning system. The inside surface of the C ontainment Structure is lined with a carbon steel liner to ensure a high degree of leak tightness during operating and

accident conditions.

The concrete building is required for structural integrity of the C ontainment Structure under DBA conditions. The steel liner and its penetrations establish the leakage limiting

boundary of the C ontainment Structure. Maintaining the C ontainment Structure OPERABLE limits the leakage of fission product radioactivity from the C ontainment Structure to the environment.

Surveillance Requirement (SR) 3.6.1.1 leakage rate requirements comply with Reference 1

, as modified by approved exemptions.

The isolation devices for the penetrations in the containment boundary are a part of the containment leak tight barrier. To

maintain this leak tight barrier:

a. All penetrations required to be closed during accident conditions are either:

1. capable of being closed by an OPERABLE automatic containment isolation system, or Containment B 3.6.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.1-2 Revision 41

2. closed by manual valves, blind flanges, or de-activated automatic valves secured in their closed positions, except as provided in Limiting Condition

for Operation (LCO) 3.6.3;

b. Each air lock is OPERABLE, except as provided in LCO 3.6.2;

c. The equipment hatch is closed and sealed.

APPLICABLE The safety design basis for the Containment Structure is SAFETY ANALYSES that the Containment Structure must withstand the pressures and temperatures of the limiting DBA without exceeding the design

leakage rate.

The DBAs that result in a release of radioactive material within

Containment Structure are a loss of coolant accident (LOCA), a

main steam line break (SLB), and a control element assembly (CEA) ejection accident (Reference 2, Chapter 14). In the analysis

of each of these accidents, it is assumed that Containment

Structure is OPERABLE, such that release of fission products to

the environment is controlled by the rate of Containment Structure leakage. The Containment Structure was designed with

an allowable leakage rate of 0.

16% of containment air weight per day (Reference 2, Chapter 5). This leakage rate is defined in Reference 1, as L a: the maximum allowable containment leakage rate at the calculated maximum peak containment pressure (P a) of 49.4 psig, which results from the limiting design basis LOCA (Reference 2, Chapter 14).

Satisfactory leakage rate test results are a requirement for the

establishment of Containment Structure OPERABILITY.

The Containment Structure satisfies 10 CFR 50.36(c)(2)(ii), Criterion 3.

LCO Containment OPERABILITY is maintained by limiting leakage to 1.0 L a (276,8 00 SCCM), except prior to the first startup after performing a required Containment Leakage Rate Testing Program leakage test. At this time the applicable leakage limits must

be met.

Containment B 3.6.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.1-3 Revision 2 Compliance with this LCO will ensure a containment configuration, including an equipment hatch, that is

structurally sound and that will limit leakage to those leakage

rates assumed in the safety analysis.

Individual leakage rates specified for the containment air lock (LCO 3.6.2) are not specifically part of the acceptance criteria of Reference 1. Therefore, leakage rates exceeding these individual limits only result in the C ontainment Structure being inoperable when the leakage results in exceeding the overall acceptance criteria of 1.0 L

a. APPLICABILITY In MODE s 1, 2, 3, and 4, a DBA could cause a release of radioactive material into the C ontainment Structure. In MODE s 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of

these MODE

s. Therefore, the C ontainment Structure is not required to be OPERABLE in MODE 5 to prevent leakage of

radioactive material from the C ontainment Structure. The requirements for the C ontainment Structure during MODE 6 are addressed in LCO 3.9.3

.

ACTIONS A.1 In the event the C ontainment Structure is inoperable, the C ontainment Structure must be restored to OPERABLE status within one hour. The one hour Completion Time provides a period of time to correct the problem commensurate with the importance of

maintaining the C ontainment Structure during MODE s 1, 2, 3, and 4. This time period also ensures that the probability of

an accident (requiring C ontainment OPERABILITY) occurring during periods when the C ontainment Structure is inoperable is minimal.

B.1 and B.2 If the C ontainment Structure cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to 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 to 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 plant conditions from full Containment B 3.6.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.1-4 Revision 41 power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.1.1 REQUIREMENTS Maintaining the Containment Structure OPERABLE requires compliance with the visual examinations and leakage rate test requirements of the Containment Leakage Rate Testing Program.

Failure to meet leakage limits specified in LCO 3.6.2 and LCO 3.6.3 does not invalidate the acceptability of these overall

leakage determinations unless their contribution to overall Type A, B, and C leakage causes that to exceed limits. As left

leakage, prior to the first startup after performing a required

Containment Leakage Rate Testing Program, is required to be 0.6 L a (166,080 SCCM) for combined Type B and C leakage and 0.75 L a (207,6 00 SCCM) for overall Type A leakage. At all other times between required leakage rat e tests, the acceptance criteria is based on an overall Type A leakage limit of 1.0 L a. At 1.0 L a , the offsite dose consequences are bounded by the assumptions of the safety analysis. Surveillance Requirement Frequencies are as required by Containment Leakage

Rate Testing Program. These periodic testing requirements

verify that the containment leakage rate does not exceed the

leakage rate assumed in the safety analysis.

SR 3.6.1.2 For ungrouted, post-tensioned tendons, t his SR ensures that the structural integrity of the Containment Structure will be

maintained in accordance with the provisions of the Concrete

Containment Tendon Surveillance Program. Testing and

Frequency are consistent with the recommendations of Reference 3.

REFERENCES 1. 10 CFR Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors" Option B, "Performance-Based Requirements" 2. Updated Final Safety Analysis Report (UFSAR)

Containment B 3.6.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.1-5 Revision 41

3. American Society of Mechanical Engineers Boiler and Pressure Vessel Code, 1992 Edition through the 1992 Addenda,Section XI, Subsection IWL, "Requirements for

Class CC Concrete Components of Light-Water Cooled Power Plants" as modified and amended by 10 CFR 50.55a

Containment Air Locks B 3.6.2 B 3.6 CONTAINMENT SYSTEMS B 3.6.2 Containment Air Locks

BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.2-1 Revision 2 BACKGROUND Containment air locks form part of the containment pressure boundary and provide a means for personnel access during all MODE s of operation.

Each air lock is nominally a right circular cylinder, 9 feet-9 inches in diameter for the personnel air lock and 5 feet-9 inches in diameter for the emergency air lock, with

a door at each end. The doors are interlocked to prevent

simultaneous opening. During periods when the C ontainment Structure is not required to be OPERABLE, the door interlock mechanism may be disabled, allowing both doors of an air

lock to remain open for extended periods when frequent

C ontainment Structure entry is necessary. Each air lock door has been designed and tested to certify its ability to

withstand a pressure in excess of the maximum expected

pressure following a DBA in the C ontainment Structure. As such, closure of a single door supports the C ontainment Structure OPERABILITY. Each of the doors contains double gasketed seals and local leakage rate testing capability to

ensure pressure integrity. To effect a leak tight seal, the

air lock design uses pressure seated doors (i.e., an

increase in containment internal pressure results in

increased sealing force on each door).

Each personnel air lock is provided with an alarm in the

C ontrol R oom that actuates when either door or equalizing valve for a personnel air lock is opened. The alarm senses

door position from a limit switch located on each door and

equalizing valve.

The containment air locks form part of the containment

pressure boundary. As such, air lock integrity and leak tightness is essential for maintaining the containment leakage rate within limit in the event of a DBA. Not

maintaining air lock integrity or leak tightness may result

in a leakage rate in excess of that assumed in the unit safety analysis.

Containment Air Locks B 3.6.2 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.2-2 Revision 41 APPLICABLE The DBAs that result in a release of radioactive material SAFETY ANALYSES within the Containment Structure are a LOCA, a main SLB, and a CEA ejection accident (Reference 1, Chapter 14). In the

analysis of each of these accidents, it is assumed that the

Containment Structure is OPERABLE such that release of

fission products to the environment is controlled by the rate of containment leakage. The Containment Structure is designed with an allowable leakage rate of 0.

16% of containment air weight per day (Reference 1, Chapter 5).

This leakage rate is defined in 10 CFR Part 50, Appendix J, Option B, as the maximum allowable containment leakage rate

at the calculated peak containment internal pressure, P a (49.4 psig), following a design basis LOCA. This allowable

leakage rate forms the basis for the acceptance criteria

imposed on the SRs associated with the air lock.

The containment air locks satisfy 10 CFR 50.36(c)(2)(ii), Criterion 3.

LCO Each containment air lock forms part of the containment

pressure boundary. As part of the containment pressure

boundary, the air lock safety function is related to control

of the containment leakage rate resulting from a DBA. Thus, each air lock's structural integrity and leak tightness, are

essential to the successful mitigation of such an event.

Each air lock is required to be OPERABLE. For the air lock to be considered OPERABLE, the air lock interlock mechanism

must be OPERABLE, the air lock must be in compliance with

the Type B air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door of an air lock to be opened at one time. This

provision ensures that a gross breach of the Containment

Structure does not exist when the Containment Structure is

required to be OPERABLE. Closure of a single door in each

air lock is sufficient to provide a leak tight barrier

following postulated events. Nevertheless, both doors are

kept closed when the air lock is not being used for normal entry into or exit from the Containment Structure.

APPLICABILITY In MODEs 1, 2, 3, and 4, a DBA could cause a release of radioactive material to the containment atmosphere. In

MODEs 5 and 6, the probability and consequences of these Containment Air Locks B 3.6.2 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.2-3 Revision 2 events are reduced due to the pressure and temperature limitations of these MODE

s. Therefore, the containment air locks are not required in MODE 5 to prevent leakage of

radioactive material from the C ontainment Structure. The requirements for the containment air locks during MODE 6 are

addressed in LCO 3.9.3

. ACTIONS The ACTIONS are modified by a Note that allows entry and exit to perform repairs on the affected air lock component.

If the outer door is inoperable, then it may be easily

accessed for most repairs. It is preferred that the air

lock be accessed from inside primary containment by entering

through the other OPERABLE air lock. However, if this is

not practicable, or if repairs on either door must be

performed from the barrel side of the door then it is

permissible to enter the air lock through the OPERABLE door, which means there is a short time during which the

containment boundary is not intact (during access through

the OPERABLE door). The ability to open the OPERABLE door, even if it means the containment boundary is temporarily not

intact, is acceptable because of the low probability of an

event that could pressurize the C ontainment Structure during the short time in which the OPERABLE door is expected to be

open. After each entry and exit, the OPERABLE door must be

immediately closed. If as low as reasonably achievable (ALARA) conditions permit, entry and exit should be via an OPERABLE air lock.

A second Note has been added to provide clarification that, for this LCO, separate Condition entry is allowed for each

air lock. This is acceptable, since the Required Actions

for each Condition provide appropriate compensatory actions

for each inoperable air lock. Complying with the Required

Actions may allow for continued operation, and a subsequent inoperable air lock is governed by subsequent Condition entry and application of associated Required Actions. A

third Note has been included that requires entry into the

applicable Conditions and Required Actions of LCO 3.6.1

, when leakage results in exceeding the overall containment

leakage limit.

Containment Air Locks B 3.6.2 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.2-4 Revision 2 A.1, A.2, and A.3 With one air lock door inoperable in one or more containment air locks, the OPERABLE door must be verified closed (Required Action A.1) in each affected containment air lock.

This ensures that a leak tight containment barrier is

maintained by the use of an OPERABLE air lock door. This action must be completed within one hour. This specified time period is consistent with the ACTIONS of LCO 3.6.1, which requires the C ontainment Structure be restored to OPERABLE status within one hour.

In addition, the affected air lock penetration must be

isolated by locking closed an OPERABLE air lock door within

the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time is

considered reasonable for locking the OPERABLE air lock

door, considering the OPERABLE door of the affected air lock

is being maintained closed.

Required Action A.3 verifies that an air lock with an inoperable door has been isolated by the use of a locked and

closed OPERABLE air lock door. This ensures that an

acceptable containment leakage boundary is maintained. The

Completion Time of once per 31 days is based on engineering

judgment and is considered adequate in view of the low

likelihood of a locked door being mispositioned and other

administrative controls. Required Action A.3 is modified by

a Note that applies to air lock doors located in high

radiation areas and allows these doors to be verified locked

closed by use of administrative means. Allowing

verification by administrative means is considered

acceptable, since access to these areas is typically

restricted. Therefore, the probability of misalignment of

the door, once it has been verified to be in the proper

position, is small.

The Required Actions have been modified by two Notes.

Note 1 ensures that only the Required Actions and associated

Completion Times of Condition C are required if both doors

in the same air lock are inoperable. With both doors in the

same air lock inoperable, an OPERABLE door is not available

to be closed. Required Actions C.1 and C.2 are the

appropriate remedial actions. The exception to Note 1 does Containment Air Locks B 3.6.2 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.2-5 Revision 2 not affect tracking the Completion Time from the initial entry into Condition A; only the requirement to comply with

the Required Actions. Note 2 allows use of the air lock for

entry and exit for seven days under administrative controls if both air locks have an inoperable door. This seven day restriction begins when the second air lock is discovered inoperable. Containment entry may be required to perform Technical Specifications Surveillances and Required Actions, as well as other activities on equipment inside the C ontainment Structure that are required by T echnical S pecifications or activities on equipment that support T echnical S pecifications-required equipment. This Note is not intended to preclude performing other activities (i.e., non-T echnical S pecifications-required activities) if the C ontainment Structure was entered, using the inoperable air lock, to perform an allowed activity listed above. This

allowance is acceptable due to the low probability of an

event that could pressurize the C ontainment Structure during the short time that the OPERABLE door is expected to be

open.

B.1, B.2, and B.3 With an air lock interlock mechanism is inoperable in one or more air locks, the Required Actions and associated

Completion Times are consistent with those specified in

Condition A.

The Required Actions have been modified by two Notes.

Note 1 ensures that only the Required Actions and associated

Completion Times of Condition C are required if both doors

in the same air lock are inoperable. With both doors in the

same air lock inoperable, an OPERABLE door is not available

to be closed. Required Actions C.1 and C.2 are the

appropriate remedial actions. Note 2 allows entry into and exit from the C ontainment Structure under the control of a dedicated individual stationed at the air lock

, to ensure that only one door is opened at a time (i.e., the individual performs the function of the interlock).

Required Action B.3 is modified by a Note that applies to

air lock doors located in high radiation areas and allows

these doors to be verified locked closed by use of Containment Air Locks B 3.6.2 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.2-6 Revision 2 administrative means. Allowing verification by administrative means is considered acceptable, since access

to these areas is typically restricted. Therefore, the

probability of misalignment of the door, once it has been

verified to be in the proper position, is small.

C.1, C.2, and C.3 With one or more air locks inoperable for reasons other than those described in Condition s A or B, Required Action C.1 requires action to be initiated immediately to evaluate

previous combined leakage rates using current air lock test

results. An evaluation is acceptable since it is overly

conservative to immediately declare the C ontainment Structure inoperable if both doors in an air lock have failed a seal test or if the overall air lock leakage is not

within limits. In many instances (e.g., only one seal per

door has failed), the C ontainment Structure remains OPERABLE, yet only one hour (per LCO 3.6.1) would be provided to restore the air lock door to OPERABLE status

prior to requiring a plant shutdown. In addition, even with

both doors failing the seal test, the overall containment

leakage rate can still be within limits.

Required Action C.2 requires that one door in the affected

containment air lock must be verified to be closed. This

action must be completed within the one hour Completion Time. This specified time period is consistent with the

ACTIONS of LCO 3.6.1, which requires that the C ontainment Structure be restored to OPERABLE status within one hour.

Additionally, the affected air lock(s) must be restored to

OPERABLE status within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time. The

specified time period is considered reasonable for restoring

an inoperable air lock to OPERABLE status, assuming that at least one door is maintained closed in each affected air lock.

D.1 and D.2 If the inoperable containment air lock cannot be restored to

OPERABLE status within the required Completion Time, the

plant must be brought to a MODE in which the LCO does not

apply. To achieve this status, the plant must be brought to Containment Air Locks B 3.6.2 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.2-7 Revision 2 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 to 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 plant

conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.2.1 REQUIREMENTS

Maintaining containment air locks OPERABLE requires

compliance with the leakage rate test requirements of the

Containment Leakage Rate Testing Program. This SR reflects

the leakage rate testing requirements with regard to air

lock leakage (Type B leakage tests). The acceptance

criteria were established during initial air lock and the C ontainment Structure OPERABILITY testing. The periodic testing requirements verify that the air lock leakage does

not exceed the allowed fraction of the overall containment

leakage rate. The Frequency is required by the Containment

Leakage Rate Testing Program.

The SR has been modified by two Notes. Note 1 states that

an inoperable air lock door does not invalidate the previous

successful performance of the overall air lock leakage test.

This is considered reasonable since either air lock door is

capable of providing a fission product barrier in the event

of a DBA. Note 2 has been added to this SR requiring the

results to be evaluated against the acceptance criteria of

which is applicable to SR 3.6.1.1. This ensures that air

lock leakage is properly accounted for in determining the

combined Type s B and C containment leakage rate.

SR 3.6.2.2 The air lock interlock is designed to prevent simultaneous

opening of both doors in a single air lock. Since both the

inner and outer doors of an air lock are designed to

withstand the maximum expected post

-accident containment pressure, closure of either door will support the C ontainment Structure OPERABILITY. Thus, the door interlock feature supports the C ontainment Support OPERABILITY while the air lock is being used for personnel transit into and

out of the C ontainment Structure. Periodic testing of this interlock demonstrates that the interlock will function as Containment Air Locks B 3.6.2 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.2-8 Revision 2 designed and that simultaneous opening of the inner and outer doors will not inadvertently occur. Due to the purely

mechanical nature of this interlock, and given that the

interlock mechanism is not normally challenged when the air

lock is used for entry and exit (procedures require strict

adherence to single door opening), this test is only required to be performed every 24 months. The 24 month Frequency is based on the need to perform this s urveillance test under the conditions that apply during a plant outage and the potential for loss of the C ontainment Structure OPERABILITY if the s urveillance test w as performed with the reactor at power. The 24 month Frequency for the interlock

is justified based on generic operating experience. The

24 month Frequency is based on engineering judgment and is

considered adequate given that the interlock is not challenged during use of the air lock.

REFERENCES

1. UFSAR Containment Isolation Valves B 3.6.3 B 3.6 CONTAINMENT SYSTEMS B 3.6.3 Containment Isolation Valves

BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.3-1 Revision 2 BACKGROUND The containment isolation valves form part of the containment pressure boundary

. They provide a means for fluid penetrations (not serving accident consequence limiting systems

) to be provided with two isolation barriers that are closed on an automatic isolation signal. These isolation devices are either passive or active (automatic).

Manual valves, de-activated automatic valves secured in

their closed position (including check valves with flow

through the valve secured), blind flanges, and closed

systems are considered passive devices. Check valves, or

other automatic valves designed to close without operator

action following an accident, are considered active devices.

Two barriers in series are provided for each penetration so

that no single credible failure or malfunction of an active

component can result in a loss of isolation or leakage that

exceeds limits assumed in the safety analysis. One of these

barriers may be a closed system.

Containment isolation occurs upon receipt of a high containment pressure signal. The containment isolation

signal closes automatic containment isolation valves in fluid penetrations

, not required for operation of Engineered Safety Feature (ESF) systems , in order to prevent leakage of radioactive material. Upon actuation of safety injection, automatic containment isolation valves also isolate systems

not required for the C ontainment Structure or Reactor Coolant System (RCS) heat removal. Other penetrations are

isolated by the use of valves in the closed position or

blind flanges. As a result, the containment isolation

valves (and blind flanges) help ensure that the containment

atmosphere will be isolated in the event of a release of

radioactive material to containment atmosphere from the RCS following a DBA. The OPERABILITY requirements for containment isolation

valves help ensure that the C ontainment Structure is isolated within the time limits assumed in the safety

analysis. Therefore, the OPERABILITY requirements provide

assurance that the C ontainment Structure function assumed in the accident analysis will be maintained.

Containment Isolation Valves B 3.6.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.3-2 Revision 32 APPLICABLE The containment isolation valve LCO was derived from the SAFETY ANALYSES assumptions related to minimizing the loss of reactor coolant inventory and establishing the containment boundary

during major accidents. As part of the containment

boundary, containment isolation valve OPERABILITY supports leak tightness of the Containment Structure. Therefore, the safety analysis of any event requiring isolation of the

Containment Structure is applicable to this LCO.

The DBAs that result in a release of radioactive material

within the Containment Structure are a LOCA, a main SLB, and

a CEA ejection accident. In the analysis for each of these

accidents, it is assumed that containment isolation valves

are either closed or function to close within the required

isolation time following event initiation. This ensures

that potential paths to the environment through containment

isolation valves (including containment purge valves) are

minimized. The safety analysis assumes that the purge

valves are closed at event initiation.

The DBA analysis assumes that, within 60 seconds after the accident, isolation of the Containment Structure is complete

and leakage terminated except for the design leakage rate, L a. The containment isolation total response time of 60 seconds includes signal delay, diesel generator startup (for loss of offsite power), and containment isolation valve

stroke times.

The containment isolation valves satisfy 10 CFR 50.36(c)(2)(ii), Criterion 3.

LCO Containment isolation valves form a part of the Containment

Structure boundary. The containment isolation valve safety

function is related to minimizing the loss of reactor

coolant inventory and establishing the Containment Structure

boundary during a DBA.

The valves covered by this LCO are listed in Reference 1.

The automatic power operated isolation valves are required to have isolation times within limits and to actuate on an automatic isolation signal. The se valves are listed with their associated stroke times in Reference

2.

Containment Isolation Valves B 3.6.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.3-3 Revision 2 The normally closed isolation valves are considered OPERABLE

when manual valves are closed, automatic valves are

de-activated and secured in their closed position, blind

flanges are in place, and closed systems are intact. These

passive isolation valves or devices are those listed in Reference 1.

This LCO provides assurance that the containment isolation

valves will perform their designed safety functions to

minimize the loss of reactor coolant inventory and establish the C ontainment Structure boundary during accidents.

APPLICABILITY In MODE s 1, 2, 3, and 4, a DBA could cause a release of radioactive material to the C ontainment Structure. In MODE s 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature

limitations of these MODE

s. Therefore, the containment isolation valves are not required to be OPERABLE in MODE 5.

The requirements for containment isolation valves during MODE 6 are addressed in LCO 3.9.3

. ACTIONS The ACTIONS are modified by a Note allowing penetration flow paths to be unisolated intermittently under administrative

controls. These administrative controls consist of stationing a dedicated operator at the valve controls who is in continuous communication with the C ontrol R oom. In this way, the penetration can be rapidly isolated when a need for containment isolation is indicated.

A second Note has been added to provide clarification that, for this LCO, separate Condition entry is allowed for each

penetration flow path. This is acceptable, since the

Required Actions for each Condition provide appropriate

compensatory actions for each inoperable containment

isolation valve. Complying with the Required Actions may

allow for continued operation, and subsequent inoperable

containment isolation valves are governed by subsequent

Condition entry and application of associated Required

Actions.

The ACTIONS are further modified by a third Note, which ensures that appropriate remedial actions are taken, if Containment Isolation Valves B 3.6.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.3-4 Revision 2 necessary, if the affected systems are rendered inoperable by an inoperable containment isolation valve.

The fourth Note has been added that requires entry into the

applicable Conditions and Required Actions of LCO 3.6.1

, when leakage results in exceeding the overall containment leakage limit.

The fifth Note allows the shutdown cooling isolation valves to be opened when RCS temperature is 300 F to establish shutdown cooling flow. This Note is required for Operation in MODE 4 to allow shutdown cooling to be established.

A.1 and A.2 In the event one containment isolation valve in one or more

penetration flow paths is inoperable, the affected

penetration flow path must be isolated. The method of

isolation must include the use of at least one isolation

barrier that cannot be adversely affected by a single active

failure. Isolation barriers that meet this criterion are a

closed and de-activated automatic containment isolation

valve, a closed manual valve, a blind flange, and a check

valve with flow through the valve secured. For penetrations

isolated in accordance with Required Action A.1, the device

used to isolate the penetration should be the closest

available one to the C ontainment Structure. Required Action A.1 must be completed within the four hour Completion Time. The four hour Completion Time is reasonable, considering the time required to isolate the penetration and

the relative importance of supporting the C ontainment Structure OPERABILITY during MODE s 1, 2, 3, and 4.

For affected penetration flow paths that cannot be restored

to OPERABLE status within the four hour Completion Time and that have been isolated in accordance with Required Action A.1, the affected penetration flow paths must be verified to be isolated on a periodic basis. This is

necessary to ensure that containment penetrations required

to be isolated following an accident and no longer capable

of being automatically isolated

, will be in the isolation position should an event occur. This Required Action does

not require any testing or device manipulation. Rather, it Containment Isolation Valves B 3.6.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.3-5 Revision 2 involves verification, through a system walkdown, that those isolation devices outside the C ontainment Structure and capable of being mispositioned are in the correct position.

The Completion Time of "once per 31 days for isolation

devices outside C ontainment" is appropriate considering the fact that the devices are operated under administrative controls and the probability of their misalignment is low.

For the isolation devices inside the C ontainment Structure , the time period specified as "prior to entering MODE 4 from

MODE 5 if not performed within the previous 92 days" is

based on engineering judgment and is considered reasonable

in view of the inaccessibility of the isolation devices and

other administrative controls that will ensure that

isolation device misalignment is an unlikely possibility.

Condition A has been modified by a Note indicating that this Condition is only applicable to those penetration flow paths

with two containment isolation valves and not a closed

system. For penetration flow paths with one or more

containment isolation valves and a closed system, Condition C provides appropriate actions.

Required Action A.2 is modified by a Note that applies to isolation devices located in high radiation areas and allows

these devices to be verified closed by use of administrative

means. Allowing verification by administrative means is

considered acceptable, since access to these areas is

typically restricted. Therefore, the probability of

misalignment of these devices, once they have been verified

to be in the proper position, is small.

B.1 With two containment isolation valves in one or more

penetration flow paths inoperable, the affected penetration flow path must be isolated within one hour. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active

failure. Isolation barriers that meet this criterion are a

closed and de-activated automatic valve, a closed manual

valve, and a blind flange. The one hour Completion Time is consistent with the ACTIONS of LCO 3.6.1. In the event the

affected penetration is isolated in accordance with Required Containment Isolation Valves B 3.6.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.3-6 Revision 2 Action B.1, the affected penetration must be verified to be isolated on a periodic basis per Required Action A.2, which

remains in effect. This periodic verification is necessary

to assure leak tightness of the C ontainment Structure and that penetrations requiring isolation following an accident

are isolated. The Completion Time of once per 31 days for verifying each affected penetration flow path is isolated

, is appropriate

, considering the fact that the valves are operated under administrative controls and the probability of their misalignment is low.

Condition B is modified by a Note indicating this Condition

is only applicable to penetration flow paths with two

containment isolation valves. Condition A of this LCO

addresses the condition of one containment isolation valve

inoperable in this type of penetration flow path.

C.1 and C.2 With one or more containment isolation valves inoperable in one or more penetration flow paths , the inoperable valves must be restored to OPERABLE status or the affected

penetration flow path must be isolated. The method of

isolation must include the use of at least one isolation

barrier that cannot be adversely affected by a single active

failure. Isolation barriers that meet this criterion are a

closed and de-activated automatic valve, a closed manual

valve, and a blind flange. A check valve may not be used to

isolate the affected penetration. Required Action C.1 must

be completed within the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time. The

specified time period is reasonable, considering the

relative stability of the closed system (hence, reliability)

to act as a penetration isolation boundary and the relative

importance of supporting the C ontainment Structure OPERABILITY during MODE s 1, 2, 3, and 4. In the event the affected penetration is isolated in accordance with Required

Action C.1, the affected penetration flow path must be

verified to be isolated on a periodic basis. This is

necessary to assure leak tightness of the C ontainment Structure and that containment penetrations requiring isolation following an accident are isolated. The

Completion Time of once per 31 days for verifying that each

affected penetration flow path is isolated

, is appropriate Containment Isolation Valves B 3.6.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.3-7 Revision 32 considering the valves are operated under administrative controls and the probability of their misalignment is low.

Condition C is modified by a Note indicating that this

Condition is only applicable to those penetration flow paths

with one or more containment isolation valves and a closed system. This Note is necessary since this Condition is written to specifically address those penetration flow paths

in a closed system. Containment isolation valves and their

associated penetration numbers are given in Reference 1. The penetrations with closed systems are listed below.

Penetration No.

Function 1B Containment Vent Header to Waste Gas 16 Component Cooling Water Inlet 17A Steam Generator Surface Blowdown 17B Steam Generator Surface Blowdown 18 Component Cooling Water Outlet 19A Instrument Air 20A Nitrogen Supply 20B Nitrogen Supply 20C Nitrogen Supply 21 Auxiliary Feedwater 22 Auxiliary Feedwater 23 Reactor Coolant Drain Tank Drains 24 Oxygen Sample Line 25 Service Water Inlet 26 Service Water Inlet 27 Service Water Inlet 28 Service Water Inlet 29 Service Water Return 30 Service Water Return 31 Service Water Return 32 Service Water Return 33 Main Feedwater 34 Main Feedwater 35 Main Steam 36 Main Steam 38 Demineralized Water 43A Steam Generator Bottom Blowdown 43B Steam Generator Bottom Blowdown 44 Fire Protection Containment Isolation Valves B 3.6.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.3-8 Revision 32 Required Action C.2 is modified by a Note that applies to

valves and blind flanges, located in high radiation areas, and allows these devices to be verified closed by use of

administrative means. Allowing verification by

administrative means is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment of these valves, once they have

been verified to be in the proper position, is small.

D.1 and D.2 If the Required Actions and associated Completion Times are

not met, the plant must be brought to a MODE in which the

LCO does not apply. To achieve this status, the plant must

be brought to 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 to 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 plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.3.1 REQUIREMENTS

This SR ensures that the containment vent valves are closed

as required, or, if open, open for an allowable reason. If

a containment vent valve is open in violation of this SR, the valve is considered inoperable. If the inoperable valve

is not otherwise known to have excessive leakage when

closed, it is not considered to have leakage outside of

limits. The SR is not required to be met when the

containment vent valves are open for pressure control, ALARA or air quality considerations for personnel entry, or for surveillance tests that require the valves to be open. The

containment vent valves are capable of closing in the

environment, following a LOCA. Therefore, these valves are

allowed to be open for limited periods of time. The 31 day

Frequency is consistent with other containment isolation

valve requirements discussed in SR 3.6.3.2.

SR 3.6.3.2 This SR requires verification that each containment

isolation manual valve and blind flange located outside the

Containment Structure, and not locked, sealed, or otherwise Containment Isolation Valves B 3.6.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.3-9 Revision 32 secured, and required to be closed during accident conditions is closed. The SR helps to ensure that post-

accident leakage of radioactive fluids or gases outside the

containment boundary is within design limits. This SR does

not require any testing or valve manipulation. Rather, it

involves verification, through a system walkdown, that those containment isolation valves outside the Containment Structure and capable of being mispositioned are in the

correct position. Since verification of valve position for

containment isolation valves outside the Containment

Structure is relatively easy, the 31 day Frequency is based

on engineering judgment, and was chosen to provide added

assurance of the correct positions. Containment isolation

valves that are open under administrative controls are not

required to meet the SR during the time the valves are open.

This SR does not apply to valves that are locked, sealed, or

otherwise secured in the closed position, since these were

verified to be in the correct position upon locking, sealing, or securing.

The Note applies to valves and blind flanges located in high radiation areas and allows these devices to be verified

closed by use of administrative means. Allowing

verification by administrative means is considered

acceptable, since access to these areas is typically

restricted during MODEs 1, 2, 3, 4 and for ALARA reasons.

Therefore, the probability of misalignment of these

containment isolation valves, once they have been verified

to be in the proper position, is small.

SR 3.6.3.3 This SR requires verification that each containment

isolation manual valve and blind flange located inside the

Containment Structure, and not locked, sealed, or otherwise secured, and required to be closed during accident conditions is closed. The SR helps to ensure that post-

accident leakage of radioactive fluids or gases outside the

containment boundary is within design limits. For

containment isolation valves inside the Containment

Structure, the Frequency of "prior to entering MODE 4 from

MODE 5 if not performed within the previous 92 days" is

appropriate, since these containment isolation valves are

operated under administrative controls and the probability Containment Isolation Valves B 3.6.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.3-10 Revision 32 of their misalignment is low. Containment isolation valves that are open under administrative controls are not required

to meet the SR during the time that they are open. This SR

does not apply to valves that are locked, sealed, or

otherwise secured in the closed position, since these were

verified to be in the correct position upon locking, sealing, or securing.

The Note allows valves and blind flanges located in high

radiation areas to be verified closed by use of

administrative means. Allowing verification by

administrative means is considered acceptable, since access

to these areas is typically restricted during MODEs 1, 2, and 3 for ALARA reasons. Therefore, the probability of

misalignment of these containment isolation valves, once

they have been verified to be in their proper position, is

small.

SR 3.6.3.4 Verifying that the isolation time of each automatic power

operated containment isolation valve is within limits is

required to demonstrate OPERABILITY. The isolation time

test, ensures the valve will isolate in a time period less

than or equal to that assumed in the safety analysis. The

isolation time and Frequency of this SR are in accordance

with the Inservice Testing Program.

The isolation time limits are contained in Reference 2.

SR 3.6.3.5 Automatic containment isolation valves close on an isolation

signal [containment isolation signal Channels A or B, or

safety injection actuation signal (SIAS) Channels A or B] to

prevent leakage of radioactive material from the Containment

Structure following a DBA. This SR ensures each automatic containment isolation valve will actuate to its isolation position on a containment isolation actuation signal. This

surveillance test is not required for valves that are

locked, sealed, or otherwise secured in the required

position under administrative controls. The 24 month

Frequency was developed considering it is prudent that this

SR be performed only during a unit outage, since isolation

of penetrations would eliminate cooling water flow and Containment Isolation Valves B 3.6.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.3-11 Revision 32 disrupt normal operation of many critical components.

Operating experience has shown that these components usually

pass this SR when performed on the 24 month Frequency.

Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

REFERENCES 1. UFSAR, Chapter 5, "Structures" , Figure 5-10

2. UFSAR, Chapter 5, "Structures", Table 5-3

Containment Pressure B 3.6.4 B 3.6 CONTAINMENT SYSTEMS B 3.6.4 Containment Pressure

BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.4-1 Revision 47 BACKGROUND The containment pressure is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a LOCA or main SLB. These limits also prevent

the containment pressure from exceeding the containment design negative pressure differential, with respect to the outside atmosphere in the event of the Containment Structure

being sealed during low barometric pressure and high

temperature, then being exposed to a concurrent cooling of

containment atmosphere and a barometric pressure rise.

Containment pressure is a process variable that is monitored and controlled. The containment pressure limits are derived

from the input conditions used in the containment functional

analyses and the containment structure external pressure

analysis. Should operation occur above the upper limits

coincident with a DBA, post-accident containment pressures

could exceed calculated values. Should containment closure

or integrity be set below the lower limits, the external

pressure limits may be exceeded during barometric pressure changes. APPLICABLE Containment internal pressure is an initial condition used SAFETY ANALYSES in the DBA analyses to establish the maximum peak containment internal pressure. The limiting DBA considered

for determining the maximum containment internal pressure is

the LOCA. A LOCA at 102% RATED THERMAL POWER and + 1.

0 psig initial containment pressure results in the highest

calculated internal containment pressure (P a) below the internal design pressure of 50.0 psig. The postulated DBAs

are analyzed assuming degraded containment ESF systems (i.e., assuming the loss of one ESF bus, which is the worst

case single active failure, resulting in one train of the

containment spray and one train of the containment coolers

being rendered inoperable). It is this maximum containment

pressure that is used to ensure that the licensing basis

dose limitations are met.

The initial pressure condition used in the containment analysis was 15.7 psia (1.0 psig). The LCO limit of 1.0 psig ensures that, in the event of an accident, the Containment Pressure B 3.6.4 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.4-2 Revision 47 maximum accident design pressure for the Containment Structure, 50 psig, is not exceeded. If a LOCA occurred

while the containment internal pressure was at the LCO value

of 1.0 psig, a total pressure below the design value of 50 psig would result.

Containment pressure satisfies 10 CFR 50.36(c)(2)(ii), Criterion 2.

LCO Maintaining containment pressure less than or equal to the LCO upper pressure limit ensures that, in the event of a

DBA, the resultant peak containment accident pressure will

remain below the containment design pressure. Maintaining

containment pressure greater than or equal to the LCO lower

pressure limit, ensures that the Containment Structure will

not exceed the design negative pressure differential following the inadvertent actuation of containment spray.

APPLICABILITY In MODEs 1, 2, 3, and 4, a DBA could cause a release of radioactive material to the Containment Structure. Since

maintaining containment pressure within limits is essential

to ensure initial conditions assumed in the accident

analysis are maintained, the LCO is applicable in MODEs 1, 2, 3, and 4.

In MODEs 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODEs. Therefore, maintaining

containment pressure within the limits of the LCO is not required in MODEs 5 or 6.

ACTIONS A.1 When containment pressure is not within the limits of the

LCO, containment pressure must be restored to within these

limits, within one hour. The Required Action is necessary

to return operation to within the bounds of the containment

analysis. The one hour Completion Time is consistent with

the ACTIONS of LCO 3.6.1 which requires that the Containment

Structure be restored to OPERABLE status within one hour.

Containment Pressure B 3.6.4 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.4-3 Revision 2 B.1 and B.2 If containment pressure cannot be restored to within limits within the required Completion Time, the plant must be

brought to a MODE in which the LCO does not apply. To

achieve this status, the plant must be brought to 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 to 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 plant conditions from full

power conditions in an orderly manner

, and without challenging plant systems.

SURVEILLANCE SR 3.6.4.1 REQUIREMENTS Verifying that containment pressure is within limits ensures that operation remains within the limits assumed in the

accident analysis. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency of this SR was

developed after taking into consideration operating

experience related to trending of containment pressure

variations during the applicable MODE

s. Furthermore, the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is considered adequate in view of other

indications available in the C ontrol R oom, including alarms, to alert the operator to an abnormal containment pressure condition.

REFERENCES None

Containment Air Temperature B 3.6.5 B 3.6 CONTAINMENT SYSTEMS B 3.6.5 Containment Air Temperature

BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.5-1 Revision 2 BACKGROUND The C ontainment S tructure serves to contain radioactive material that may be released from the reactor core following a DBA. The C ontainment Structure average air temperature is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a LOCA or main SLB.

The containment average air temperature limit is derived

from the input conditions used in the containment functional

analyses and the C ontainment S tructure external pressure analyses. This LCO ensures that initial conditions assumed

in the analysis of containment response to a DBA

, are not violated during unit operations. The total amount of energy

to be removed from C ontainment Structure by the c ontainment s pray and containment c ooling during post

-accident conditions is dependent on the energy released to the

C ontainment Structure due to the event, as well as the initial containment temperature and pressure. The higher

the initial temperature, the more energy that must be

removed, resulting in a higher peak containment pressure and

temperature. Exceeding containment design pressure may

result in leakage greater than that assumed in the accident

analysis (Reference 1). Operation with containment

temperature in excess of the LCO limit violates an initial condition assumed in the accident analysis.

APPLICABLE Containment average air temperature is an initial condition SAFETY ANALYSES used in the DBA analyses that establishes the containment environmental qualification operating envelope for both

pressure and temperature. The limit for containment average

air temperature ensures that operation is maintained within

the assumptions used in the DBA analysis for C ontainment.

The accident analyses and evaluations considered both LOCAs

and main SLBs for determining the maximum peak containment pressures and temperatures. The worst case LOCA generates

larger mass and energy releases than the worst case main SLB. Thus, the LOCA event bounds the main SLB event from the containment peak pressure and temperature standpoint.

The initial pre-accident temperature inside the C ontainment Structure was assumed to be 120°F (Reference 1).

Containment Air Temperature B 3.6.5 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.5-2 Revision 2 The initial containment average air temperature condition of

120°F resulted in a maximum vapor temperature described in Reference 1. The consequence of exceeding the design temperature for extended periods may be the potential for degradation of the containment structure under accident loads.

Containment average air temperature satisfies 10 CFR 50.36(c)(2)(ii), Criterion 2.

LCO During a DBA, with an initial containment average air

temperature less than or equal to the LCO temperature limit, the resultant peak accident temperature is maintained below

the containment design temperature. As a result, the

ability of the C ontainment Structure to perform its function is ensured.

APPLICABILITY In MODE s 1, 2, 3, and 4, a DBA could cause a release of radioactive material to the C ontainment Structure. In MODE s 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature

limitations of these MODE

s. Therefore, maintaining containment average air temperature within the limit is not required in MODE s 5 or 6. ACTIONS A.1 When containment average air temperature is not within the

limit of the LCO, it must be restored to within limit

, within eight hours. This Required Action is necessary to return operation to within the bounds of the containment

analysis. The eight hour Completion Time is acceptable considering the sensitivity of the analysis to variations in

this parameter and provides sufficient time to correct minor

problems.

B.1 and B.2 If the containment average air temperature cannot be

restored to within its limit

, within the required Completion Time, the plant must be brought to a MODE in which the LCO

does not apply. To achieve this status, the plant must be

brought to 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 to MODE 5 Containment Air Temperature B 3.6.5 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.5-3 Revision 3 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 plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.5.1 REQUIREMENTS

Verifying that containment average air temperature is within

the LCO limit ensures that containment operation remains

within the limit assumed for the containment analyses. In

order to determine the containment average air temperature, an arithmetic average is calculated using measurements taken

from the containment dome

[1(2)-TI-5309]

and the containment reactor cavity

[1(2)-TI-5311]

temperature indicators selected to provide a representative sample of the overall

containment atmosphere. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency of this SR is

considered acceptable based on the observed slow rates of

temperature increase within the Containment Structure as a

result of environmental heat sources (due to the large

volume of the Containment Structure). Furthermore, the

24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is considered adequate in view of other

indications available in the Control Room, including alarms, to alert the operator to an abnormal containment temperature condition.

REFERENCES 1. UFSAR, Section 14.20, "Containment Response" Containment Spray and Cooling Systems B 3.6.6 B 3.6 CONTAINMENT SYSTEMS B 3.6.6 Containment Spray and Cooling Systems

BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.6-1 Revision 2 BACKGROUND The Containment Spray and Cooling Systems provide containment atmosphere cooling to limit post-accident pressure and temperature in the Containment Structure to less than the design values. Reduction of containment pressure and the iodine removal capability of the spray, reduce the release of fission product radioactivity from the Containment Structure to the environment, in the event of a DBA, to within limits. The Containment Spray and Cooling Systems are designed to the requirements in Reference 1, Appendix 1C, Criteria, 58, 59, 60, 61, 62, 63, 64, and 65.

The Containment Spray and Cooling Systems are ESF systems.

They are designed to ensure that the heat removal capability

required during the post-accident period can be attained.

The Containment Spray and Cooling Systems provide redundant methods to limit and maintain post-accident conditions to less than the containment design values.

Containment Spray System The Containment Spray System consists of two separate trains

of equal capacity, each of sufficient capacity to supply

approximately 50% of the design cooling requirement. Each

train includes a containment spray pump, spray headers, nozzles, valves, and piping. Each train is powered from a

separate ESF bus. The refueling water tank (RWT) supplies

borated water to the containment spray during the injection

phase of operation. In the recirculation mode of operation, containment spray pump suction is transferred from the RWT

to the containment sump(s). Each spray system flow path

from the containment sump will be via an OPERABLE shutdown

cooling heat exchanger.

The Containment Spray System provides a spray of cold borated water into the upper regions of the Containment Structure to reduce containment pressure and temperature and to reduce the concentration of fission products in the containment atmosphere during a DBA. The RWT solution

temperature is an important factor in determining the heat

removal capability of the Containment Spray System during

the injection phase. In the recirculation mode of Containment Spray and Cooling Systems B 3.6.6 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.6-2 Revision 2 operation, heat is removed from the containment sump water by the shutdown cooling heat exchangers. Each train of the

Containment Spray System provides adequate spray coverage to

meet 50% of the system design requirements for containment

heat removal and 100% of the iodine removal design bases.

The Containment Spray System is actuated either automatically by a containment spray actuation signal coincident with a SIAS or manually. An automatic actuation starts the two containment spray pumps, and begins the injection phase. The containment spray header isolation valves open upon a containment spray actuation signal. A manual actuation of the Containment Spray System is

available on the main control board to begin the same

sequence. The injection phase continues until an RWT low level signal is received. The low level for the RWT generates a recirculation actuation signal that aligns valves from the containment spray pump suction to the

containment sump. The Containment Spray System in

recirculation mode maintains an equilibrium temperature

between the containment atmosphere and the recirculated sump

water. Operation of the Containment Spray System in the

recirculation mode is controlled by the operator in

accordance with the Emergency Operating Procedures.

Containment Cooling System Two trains of containment cooling, each of sufficient

capacity to supply approximately 67% of the design cooling

requirement, are provided. Two trains with two fan units

each are supplied with cooling water from a separate train

of service water cooling. Three of the four fans are

required to furnish the design cooling capacity. Air is

drawn into the coolers through the fans and discharged

throughout the Containment Structure.

In post-accident operation following a SIAS, all four Containment Cooling System fans are designed to start

automatically in slow speed. Cooling is supplied by the

service water cooled coils. The temperature of the service

water is an important factor in the heat removal capability of the fan units.

Containment Spray and Cooling Systems B 3.6.6 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.6-3 Revision 39 APPLICABLE The Containment Spray and Cooling Systems limit the SAFETY ANALYSES temperature and pressure that could be experienced following a DBA. The limiting DBAs considered relative to containment

temperature and pressure are LOCA and main SLB. The DBA, LOCA, and main SLB are analyzed using computer codes

designed to predict the resultant containment pressure and temperature transients. No DBAs are assumed to occur simultaneously or consecutively. The postulated DBAs are

analyzed with regard to containment ESF systems, assuming

the loss of one ESF bus, which is the worst case single

active failure, resulting in one train of the Containment

Spray System and one train of the Containment Cooling System

being rendered inoperable.

The analysis and evaluation show that under the worst case scenario, the highest peak containment pressure and

temperature are within the design. (See the Bases for

Specifications 3.6.4 and 3.6.5 for a detailed discussion.)

The analyses and evaluations assume a power level of 2754 MWt, one containment spray train and one containment cooling train operating, and initial (pre-accident)

conditions of 120°F and 16.5 psia. The analyses also

assumes a response time delayed initiation, in order to

provide a conservative calculation of peak containment

pressure and temperature responses.

The modeled Containment Spray System actuation from the containment analysis is based upon a response time

associated with exceeding the Containment High-High pressure

setpoint coincident with an SIAS to achieve full flow

through the containment spray nozzles. The Containment

Spray System total response time of 62.9 seconds for a main

steam line break and 70.9 seconds for a LOCA, includes

diesel generator startup (for loss of offsite power), sequencing equipment onto the emergency bus, containment spray pump startup, and spray line filling (Reference 1, Chapter 7).

The performance of the containment cooling train for post-accident conditions is given in Reference 1, Chapter 6. The

results of the analysis, is that each train can provide

approximately 67% of the required peak cooling capacity

during the post-accident condition. The train post-accident Containment Spray and Cooling Systems B 3.6.6 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.6-4 Revision 15 cooling capacity under varying containment ambient conditions, required to perform the accident analyses, is

also shown in Reference 1, Chapter 6.

The modeled Containment Cooling System actuation from the containment analysis, is based upon the unit specific response time associated with exceeding the SIAS to achieve full Containment Cooling System air and safety grade cooling

water flow.

The Containment Spray and Cooling Systems satisfy 10 CFR 50.36(c)(2)(ii), Criterion 3.

LCO During a DBA, a minimum of one containment cooling train and one containment spray train, is required to maintain the

containment peak pressure and temperature, below the design

limits (Reference 1, Chapter 6). Additionally, one

containment spray train is also required to remove iodine

from the containment atmosphere and maintain concentrations

below those assumed in the safety analysis. To ensure that

these requirements are met, two containment spray trains and

two containment cooling trains (all four coolers) must be

OPERABLE. Therefore, in the event of an accident, the

minimum requirements are met, assuming that the worst case

single active failure occurs.

Each Containment Spray System includes a spray pump, spray headers, nozzles, valves, piping, instruments, and controls

to ensure an OPERABLE flow path capable of taking suction

from the RWT upon an ESF actuation signal and automatically transferring suction to the containment sump. Each spray system flow path from the containment sump will be via an

OPERABLE shutdown cooling heat exchanger.

Each Containment Cooling System includes cooling coils, dampers, fans, instruments, and controls to ensure an OPERABLE flow path.

APPLICABILITY In MODEs 1, 2, and 3, a DBA could cause a release of radioactive material to the Containment Structure and an

increase in containment pressure and temperature, requiring

the operation of the containment spray trains and

containment cooling trains.

Containment Spray and Cooling Systems B 3.6.6 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.6-5 Revision 48 The Containment Spray System is only required to be OPERABLE in MODE 3 with pressurizer pressure 1750 psia.

In MODE 3 with pressurizer pressure

< 1750 psia, and in MODEs 4, 5, and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODEs. Thus, the Containment Spray

System is not required to be OPERABLE in MODE 3 with pressurizer pressure

< 1750 psia, and the Containment Spray and Cooling Systems are not required to be OPERABLE in MODEs 4, 5, and 6.

ACTIONS A.1 With one containment spray train inoperable, the inoperable

containment spray train 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 />. In this Condition, the remaining OPERABLE

spray and cooling trains are adequate to perform the iodine

removal and containment cooling functions. 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 redundant heat

removal capability afforded by the Containment Spray System, reasonable time for repairs, and the low probability of a

DBA occurring during this period.

B.1 and B.2 If the inoperable containment spray train cannot be restored

to OPERABLE status within the required Completion Time, the

plant must be brought to a MODE in which the LCO does not

apply. To achieve this status, the plant must be brought to

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 to MODE 3 with pressurizer pressure

< 1750 psia within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Time of six hours is reasonable, based on operating experience, to reach MODE 3 from full power

conditions in an orderly manner, and without challenging

plant systems. The extended interval to reach MODE 3 with pressurizer pressure

< 1750 psia allows additional time for the restoration of the containment spray train and is reasonable when considering that the driving force for a

release of radioactive material from the RCS is reduced in

MODE 3.

Containment Spray and Cooling Systems B 3.6.6 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.6-6 Revision 48 C.1 With one required containment cooling train inoperable, the inoperable containment cooling train must be restored to

OPERABLE status within seven days. The remaining OPERABLE

containment spray and cooling components provide iodine

removal capabilities and are capable of providing at least 100% of the heat removal needs after an accident. The seven day Completion Time was developed taking into account

the redundant heat removal capabilities afforded by

combinations of the Containment Spray and Cooling Systems, and the low probability of a DBA occurring during this

period. D.1 With two required containment cooling trains inoperable, one

of the required containment cooling trains 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 />. The remaining OPERABLE

containment spray components provide iodine removal

capabilities and are capable of providing at least 100% of

the heat removal needs after an accident. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

Completion Time was developed taking into account the

redundant heat removal capabilities afforded by combinations

of the Containment Spray and Cooling Systems, the iodine

removal function of the Containment Spray System, and the

low probability of a DBA occurring during this period.

E.1 and E.2 If the Required Actions and associated Completion Times of

Conditions C or D of this LCO are not met, the plant must be

brought to a MODE in which the LCO does not apply. To

achieve this status, the plant must be brought to 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 to 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 plant conditions from full power conditions in an orderly manner and without challenging plant systems.

F.1 With two containment spray trains or any combination of

three or more Containment Spray and Cooling Systems trains Containment Spray and Cooling Systems B 3.6.6 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.6-7 Revision 48 inoperable, the unit is in a condition outside the accident analysis. Therefore, LCO 3.0.3 must be entered immediately.

SURVEILLANCE SR 3.6.6.1 REQUIREMENTS Verifying the correct alignment for manual, power-operated, and automatic valves in the containment spray flow path

provides assurance that the proper flow paths will exist for

Containment Spray System operation. This SR does not apply

to valves that are locked, sealed, or otherwise secured in

position since these were verified to be in the correct

position prior to being secured. 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 verifying, through a system walkdown, that those valves outside the

Containment Structure and capable of potentially being

mispositioned are in the correct position.

SR 3.6.6.2 Starting each containment cooling train fan unit from the Control Room and operating it for 15 minutes ensures that all trains are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or

motor failure, or excessive vibration can be detected and

corrective action taken. The 31 day Frequency of this SR

was developed considering the known reliability of the fan

units and controls, the two train redundancy available, and

the low probability of a significant degradation of the

containment cooling train occurring between surveillances and has been shown to be acceptable through operating experience.

SR 3.6.6.3 Verifying a service water flow rate of 2000 gpm to each cooling unit when the full flow service water outlet valves

are fully open provides assurance that the design flow rate

assumed in the safety analyses will be achieved (Reference 1, Chapter 7). Also considered in selecting this

Frequency were the known reliability of the Service Water

System, the two train redundancy, and the low probability of Containment Spray and Cooling Systems B 3.6.6 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.6-8 Revision 48 a significant degradation of flow occurring between surveillance tests.

SR 3.6.6.4 Verifying that each containment spray pump's developed head

at the flow test point is greater than or equal to the required developed head ensures that spray pump performance has not degraded during the cycle. Flow and differential

pressure are normal tests of centrifugal pump performance

required by Reference 2. Since the containment spray pumps

cannot be tested with flow through the spray headers, they

are tested on recirculation flow. This test confirms one

point on the pump design curve and is indicative of overall

performance. Such inservice inspections confirm component

OPERABILITY, trend performance, and detect incipient

failures by indicating abnormal performance. The Frequency

of this SR is in accordance with the Inservice Testing

Program.

SR 3.6.6.5 and SR 3.6.6.6 These SRs verify that each automatic containment spray valve

actuates to its correct position and that each containment

spray pump starts upon receipt of an actual or simulated

actuation signal (i.e., the appropriate Engineered Safety

Feature Actuation System signal). This SR is not required

for valves that are locked, sealed, or otherwise secured in

the required position under administrative controls. The

24 month Frequency is based on the need to perform these

surveillance tests under the conditions that apply during a

plant outage and the potential for an unplanned transient if

the surveillance tests were performed with the reactor at

power. Operating experience has shown that these components

usually pass the surveillance tests when performed at the

24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

The surveillance test of containment sump isolation valves is also required by SR 3.5.2.5. A single surveillance test

may be used to satisfy both requirements.

Containment Spray and Cooling Systems B 3.6.6 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.6-9 Revision 48 SR 3.6.6.7 This SR verifies that each containment cooling train actuates upon receipt of an actual or simulated actuation

signal (i.e., the appropriate Engineered Safety Feature

Actuation System signal). The 24 month Frequency is based

on engineering judgment and has been shown to be acceptable through operating experience. See SR 3.6.6.5 and SR 3.6.6.6, above, for further discussion of the basis for

the 24 month Frequency.

SR 3.6.6.8 With the containment spray inlet valves closed and the spray

header drained of any solution, low pressure air or smoke

can be blown through check valve bonnets. Performance of

this SR demonstrates that each spray nozzle is unobstructed

and provides assurance that spray coverage of the

Containment Structure during an accident is not degraded.

Due to the passive design of the nozzle, a test after

maintenance that could result in nozzle blockage is

considered adequate. Maintenance that could result in

nozzle blockage is generally loss of foreign material

control or a flow of borated water through a nozzle. Should

either of these events occur, a supervisory evaluation will

be required to determine whether nozzle blockage is a possible result of the event.

REFERENCES 1. UFSAR

2. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section XI, "Rules for In-Service Inspection of Nuclear Power Plant Components"

IRS B 3.6.8 B 3.6 CONTAINMENT SYSTEMS B 3.6.8 Iodine Removal System (IRS)

BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.8-1 Revision 41 BACKGROUND The IRS is provided per Reference 1, Appendix 1C, Criteria 62, 63, and 64, to reduce the concentration of fission products released to the containment atmosphere

following a postulated accident. The IRS would function together with the Containment Spray and Cooling Systems following a DBA to reduce the potential release of

radioactive material, principally iodine, from the

Containment Structure to the environment.

The IRS consists of three 50% capacity separate, independent (except for power), and redundant trains. Each train

includes a moisture separator, a high efficiency particulate

air filter, an activated charcoal adsorber section for

removal of radioiodines, a fan, and instrumentation. The

moisture separators function to reduce the moisture content

of the air stream. The system initiates filtered

recirculation of the containment atmosphere following

receipt of a SIAS. The system design is described in

Reference 1, Section 6.7.

The moisture separator is included for moisture (free water) removal from the gas stream. The moisture separator is

important to the effectiveness of the charcoal adsorbers.

Three IRS trains are provided to meet the requirement for separation, independence (except for power), and redundancy.

Two trains of the IRS are powered by separate ESF buses.

The third IRS train is a swing train that can be aligned to take power from either ESFs bus.

APPLICABLE The DBAs that result in a release of radioactive iodine SAFETY ANALYSES within the Containment Structure are a LOCA, a main SLB, or a CEA ejection accident. In the analysis for each of these

accidents, it is assumed that adequate containment leak

tightness exists at event initiation to limit potential

leakage to the environment. Additionally, for the LOCA and CEA ejection event, it is assumed that the amount of radioactive iodine release is limited by reducing the iodine

concentration in the containment atmosphere.

IRS B 3.6.8 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.8-2 Revision 41 The IRS design basis is established by the consequences of the limiting DBA, which is a LOCA based Maximum Hypothetical Accident (Reference 1, Section 14.24). The accident analysis (Reference 1, Section 14.2

4) assumes that one iodine removal unit starts within 63 seconds and a second iodine removal unit is assumed to be manually started within 20 minutes. For a CEA ejection event, two iodine removal units are assumed to be manually started within 20 minutes.

The accident analysis accounts for the reduction in airborne

radioactive iodine provided by the remaining two trains of

this filtration system.

The IRS satisfies 10 CFR 50.36(c)(2)(ii), Criterion 3.

LCO Three separate, independent (except for power), and

redundant trains of the IRS are required to ensure that at

least two are available, assuming a single failure coincident with a loss of offsite power.

APPLICABILITY In MODEs 1, 2, 3, and 4, iodine is a fission product that can be released from the fuel to the reactor coolant as a

result of a DBA. The DBAs that can cause a failure of the

fuel cladding are a LOCA, main SLB, and CEA ejection

accident. Because these accidents are considered credible

accidents in MODEs 1, 2, 3, and 4, the IRS must be operable in these MODEs to ensure the reduction in iodine concentration assumed in the accident analysis.

In MODEs 5 and 6, the probability and consequences of a LOCA

are low due to the pressure and temperature limitations of

these MODEs. The IRS is not required in these MODEs to remove iodine from the containment atmosphere.

ACTIONS A.1 With one IRS train inoperable, the inoperable train must be

restored to OPERABLE status within seven days. The

components in this degraded condition are capable of

providing 100% of the iodine removal needs after a DBA. The

seven day Completion Time is based on consideration of such

factors as:

a. The availability of the OPERABLE redundant IRS train; IRS B 3.6.8 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.8-3 Revision 41

b. The fact that, even with no IRS train in operation, almost the same amount of iodine would be removed from the containment atmosphere through absorption by the

Containment Spray System; and

c. The fact that the Completion Time is adequate to make most repairs.

B.1 If two IRS trains are inoperable, one must be restored to

OPERABLE status within one hour. The one hour Completion

Time allows the swing train to be aligned to the appropriate

bus to ensure each of the two remaining trains are powered

from separate and independent buses. The one hour, also

allows time to restore one train to OPERABLE status prior to

initiating a plant shutdown. This is reasonable considering

that a plant shutdown is a plant transient.

C.1 and C.2 If the IRS train cannot be restored to OPERABLE status

within the required Completion Time, the plant must be

brought to a MODE in which the LCO does not apply. To

achieve this status, the plant must be brought to 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 to 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 plant conditions from full

power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.8.1 REQUIREMENTS Initiating each IRS train from the Control Room and operating it for 15 minutes ensures that all trains are OPERABLE and that all associated controls are functioning properly. It also ensures that motor failure can be

detected for corrective action. The 31 day Frequency was

developed considering the known reliability of fan motors

and controls, the two train redundancy available, and the

iodine removal capability of the Containment Spray System

independent of the IRS.

IRS B 3.6.8 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.6.8-4 Revision 41 SR 3.6.8.2 This SR verifies that the required IRS filter testing is performed in accordance with the Ventilation Filter Testing

Program. The IRS filter tests are in accordance with

portions of Reference 2. The Ventilation Filter Testing

Program includes testing high efficiency particulate air 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 Ventilation

Filter Testing Program.

SR 3.6.8.3 The automatic startup test verifies that both trains of

equipment start upon receipt of an actual or simulated test

signal (Engineered Safety Feature Actuation System). The

24 month Frequency is based on the need to perform this

surveillance test under the conditions that apply during a

plant outage and the potential for an unplanned transient if

the surveillance test were performed with the reactor at

power. Operating experience has shown that these components

usually pass the surveillance test when performed at the

24 month Frequency. Therefore, the Frequency was concluded

to be acceptable from a reliability standpoint.

Furthermore, the Frequency was developed considering that

the system equipment OPERABILITY is demonstrated on a 31 day Frequency by SR 3.6.8.1.

REFERENCES 1. UFSAR 2. Regulatory Guide 1.52, Revision 2, "Design, Testing, and Maintenance Criteria for Postaccident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration

and Adsorption Units of Light-Water-Cooled Nuclear Power Plants," March 1978