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Developmental Revision B - Technical Specifications Bases B 3.6 - Containment Systems
	ML100550590
Person / Time
Site:	Watts Bar
Issue date:	02/02/2010
From:	; Tennessee Valley Authority
To:	; Office of Nuclear Reactor Regulation
References
	Download: ML100550590 (6)
	v • d • e

Containment B 3.6.1 (continued)

Watts Bar - Unit 2 B 3.6-1 (developmental)

A B 3.6 CONTAINMENT SYSTEMS B 3.6.1 Containment

BASES BACKGROUND The containment is a free standing steel pressure vessel surrounded by a reinforced concrete shield building. The containment vessel, including all

its penetrations, is a low leakage steel shell designed to contain the

radioactive material that may be released from the reactor core following

a Design Basis Accident (DBA). Additionally, the containment and shield

building provide shielding from the fission products that may be present in

the containment atmosphere following accident conditions.

The containment vessel is a vertical cylindrical steel pressure vessel with

hemispherical dome and a concrete base mat with steel membrane. It is completely enclosed by a reinforced concrete shield building. An annular space exists between the walls and domes of the steel containment

vessel and the concrete shield building to provide for the collection, mixing, holdup, and controlled release of containment out leakage. Ice

condenser containments utilize an outer concrete building for shielding

and an inner steel containment for leak tightness.

Containment piping penetration assemblies provide for the passage of

process, service, sampling, and instrumentation pipelines into the

containment vessel while maintaining containment integrity. The shield

building provides shielding and allows controlled filtered release of the

annulus atmosphere under accident conditions, as well as environmental

missile protection for the containment vessel and Nuclear Steam Supply

System.

The inner steel containment and its penetrations establish the leakage

limiting boundary of the containment. Maintaining the containment

OPERABLE limits the leakage of fission product radioactivity from the

containment to the environment. SR 3.6.1.1 leakage rate requirements

comply with 10 CFR 50, Appendix J, Option B (Ref. 1), as modified by

approved exemptions.

Containment B 3.6.1 BASES (continued)

Watts Bar - Unit 2 B 3.6-2 (developmental)

A BACKGROUND (continued)

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

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

provided in LCO 3.6.3, "Containment Isolation Valves."

b. Each air lock is OPERABLE, except as provided in LCO 3.6.2, "Containment Air Locks."

c. All equipment hatches are closed.

APPLICABLE

SAFETY ANALYSES The safety design basis for the containment is that the containment must

withstand the pressures and temperatures of the limiting DBA without

exceeding the design leakage rates.

The DBAs that result in a challenge to containment OPERABILITY from

high pressures and temperatures are a loss of coolant accident (LOCA),

a steam line break (SLB), and a rod ejection accident (REA) (Ref. 2). In

addition, release of significant fission product radioactivity within

containment can occur from a LOCA or REA. In the DBA analyses, it is assumed that the containment is OPERABLE such that, for the DBAs involving release of fission product radioactivity, release to the

environment is controlled by the rate of containment leakage. The

containment was designed with an allowable leakage rate of 0.25% of

containment air weight per day (Ref. 3). This leakage rate, used in the

evaluation of offsite doses resulting from accidents, is defined in

10 CFR 50, Appendix J, Option B (Ref. 1), as L a: the maximum allowable containment leakage rate at the calculated peak containment internal

pressure (P a) related to the design basis LOCA. The allowable leakage rate represented by L a forms the basis for the acceptance criteria imposed on all containment leakage rate testing. L a is assumed to be 0.25% per day in the safety analysis at P a = 15.0 psig which bounds the calculated peak containment internal pressure resulting from the limiting

design basis LOCA (Ref. 3).

Containment B 3.6.1 BASES (continued)

Watts Bar - Unit 2 B 3.6-3 (developmental)

A APPLICABLE SAFETY ANALYSES (continued)

Satisfactory leakage rate test results are a requirement for the

establishment of containment OPERABILITY.

The containment satisfies Criterion 3 of the NRC Policy Statement.

LCO Containment OPERABILITY is maintained by limiting leakage to 1.0 L a , except prior to the first start up after performing a required Containment

Leakage Rate Testing Program leakage test. At this time, applicable

leakage limits must be met.

Compliance with this LCO will ensure a containment configuration, including equipment hatches, 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), purge valves with resilient seals, and Shield Building

containment bypass leakage (LCO 3.6.3) are not specifically part of the acceptance criteria of 10 CFR 50, Appendix J, Option B. Therefore, leakage rates exceeding these individual limits only result in the

containment being inoperable when the leakage results in exceeding the

acceptance criteria of Appendix J, Option B.

APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material into containment. 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, containment is not

required to be OPERABLE in MODE 5 to prevent leakage of radioactive

material from containment. The requirements for containment during

MODE 6 are addressed in LCO 3.9.4, "Containment Penetrations."

Containment B 3.6.1 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.6-4 (developmental)

A ACTIONS A.1 In the event containment is inoperable, containment must be restored to

OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The 1-hour Completion Time provides

a period of time to correct the problem commensurate with the

importance of maintaining containment OPERABLE during MODES 1, 2, 3, and 4. This time period also ensures that the probability of an accident (requiring containment OPERABILITY) occurring during periods when

containment is inoperable is minimal.

B.1 and B.2

If containment 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . 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

REQUIREMENTS SR 3.6.1.1

Maintaining the containment OPERABLE requires compliance with the

visual examinations and leakage rate test requirements of the

Containment Leakage Rate Testing Program. Failure to meet air lock, Shield Building containment bypass leakage path, and purge valve with

resilient seal 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 leakage test is required to be < 0.6 L a for combined Type B and C leakage and 0.75 L a for overall Type A leakage. At all other times between required leakage rate 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.

SR Frequencies are as required by the 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.

Containment B 3.6.1 BASES (continued)

Watts Bar - Unit 2 B 3.6-5 (developmental)

A REFERENCES

1. Title 10, Code of Federal Regulations, Part 50, Appendix J, Option B, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors - Performance-Based

Requirements." 2. Watts Bar FSAR, Section 15.0, "Accident Analysis." 3. Watts Bar FSAR, Section 6.2, "Containment Systems." 4. Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," September 1995.

Containment Air Locks B 3.6.2 (continued)

Watts Bar - Unit 2 B 3.6-6 (developmental)

A B 3.6 CONTAINMENT SYSTEMS B 3.6.2 Containment Air Locks

BASES BACKGROUND Containment air locks form part of the containment pressure boundary and provide a means for personnel access during all MODES of

operation.

Each air lock is nominally a right circular cylinder, 8 ft 7 inches in

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

simultaneous opening. During periods when containment 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 containment 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 Design Basis

Accident (DBA) in containment. As such, closure of a single door

supports containment 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 limit switches on both doors that

provide control room indication of door position.

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 plant safety analyses.

Containment Air Locks B 3.6.2 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.6-7 (developmental)

A APPLICABLE SAFETY ANALYSES The DBAs that result in a significant release of radioactive material within

containment are a loss of coolant accident and a rod ejection accident (Ref. 2). In the analysis of each of these accidents, it is assumed that

containment is OPERABLE such that release of fission products to the

environment is controlled by the rate of containment leakage. The

containment was designed with an allowable leakage rate (L a) of 0.25%

of containment air weight per day (Ref. 2), at the calculated peak

containment pressure of 15.0 psig. This allowable leakage rate forms the

basis for the acceptance criteria imposed on the SRs associated with the

air locks.

The containment air locks satisfy Criterion 3 of the NRC Policy

Statement.

LCO Each containment air lock forms part of the containment pressure boundary. As part of 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 containment does not exist when

containment 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 and exit from containment.

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

locks are not required in MODE 5 to prevent leakage of radioactive

material from containment. The requirements for the containment air

locks during MODE 6 are addressed in LCO 3.9.4, "Containment

Penetrations."

Containment Air Locks B 3.6.2 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.6-8 (developmental)

A 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 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

due to the low probability of an event that could pressurize the

containment 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.

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.

In the event the air lock leakage results in exceeding the overall

containment leakage rate, Note 3 directs entry into the applicable

Conditions and Required Actions of LCO 3.6.1, "Containment."

A.1, A.2, and A.3

With one air lock door in one or more containment air locks inoperable, 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 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . This specified time

period is consistent with the ACTIONS of LCO 3.6.1, which requires

containment be restored to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

In addition, the affected air lock penetration must be isolated by locking closed the OPERABLE air lock door within the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is reasonable for locking the OPERABLE

air lock door, considering the OPERABLE door of the affected air lock is

being maintained closed.

Containment Air Locks B 3.6.2 BASES (continued)

Watts Bar - Unit 2 B 3.6-9 (developmental)

A ACTIONS A.1, A.2, and A.3 (continued)

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 of Note 1 does 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 7 days under administrative controls if both air locks have an inoperable door.

This 7 day restriction begins when the second air lock is discovered

inoperable. Containment entry may be required on a periodic basis to

perform Technical Specifications (TS) Surveillances and Required

Actions, as well as other activities on equipment inside containment that

are required by TS or activities on equipment that support TS-required

equipment. This Note is not intended to preclude performing other activities (i.e., non-TS-required activities) if the containment is 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 containment during the short time that the

OPERABLE door is expected to be open.

Containment Air Locks B 3.6.2 BASES (continued)

Watts Bar - Unit 2 B 3.6-10 (developmental)

A ACTIONS (continued)

B.1, B.2, and B.3 With an air lock interlock mechanism 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 containment 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 administrativ e 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 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 containment 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), containment remains OPERABLE, yet only 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (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 within the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time.

This specified time period is consistent with the ACTIONS of LCO 3.6.1, which requires that containment be restored to OPERABLE status within

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Containment Air Locks B 3.6.2 BASES (continued)

Watts Bar - Unit 2 B 3.6-11 (developmental)

A ACTIONS C.1, C.2, and C.3 (continued)

Additionally, the affected air lock must be restored to OPERABLE status

within the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5

within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . 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

REQUIREMENTS SR 3.6.2.1

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 containment

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 requires the results of the

air lock leakage tests to be evaluated against the acceptance criteria of

the Containment Leakage Rate Testing Program, 5.7.2.19. This ensures

that air lock leakage is properly accounted for in determining the

combined Type B and C containment leakage rate.

Containment Air Locks B 3.6.2 BASES Watts Bar - Unit 2 B 3.6-12 (developmental)

A SURVEILLANCE REQUIREMENTS (continued)

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 containment

OPERABILITY. Thus, the door interlock feature supports containment

OPERABILITY while the air lock is being used for personnel transit in and

out of the containment. Periodic testing of this interlock demonstrates

that the interlock will function as 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 only challenged when the containment air lock

door is opened, this test is only required to be performed upon entering or

exiting a containment air lock but is not required more frequently than

every 184 days. The 184 day Frequency is based on engineering

judgment and is considered adequate in view of other indications of door

status available to operations personnel and because the interlock is only

disabled in MODES 5 and 6.

REFERENCES

1. Title 10, Code of Federal Regulations, Part 50, Appendix J, Option B, "Primary Reactor Containment Leakage Testing for

Water-Cooled Power Reactors - Performance-Based

Requirements." 2. Watts Bar FSAR, Section 15.0, "Accident Analysis."

Containment Isolation Valves B 3.6.3 (continued)

Watts Bar - Unit 2 B 3.6-13 (developmental)

A B 3.6 CONTAINMENT SYSTEMS B 3.6.3 Containment Isolation Valves

BASES BACKGROUND The containment isolation valves form part of the containment pressure boundary and provide a means for fluid penetrations not serving accident

consequence limiting systems to be provided with two isolation barriers

that are closed on a containment isolation signal or which are normally

closed. 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 analyses.

One of these barriers may be a closed system. These barriers (typically

containment isolation valves) make up the Containment Isolation System.

Automatic isolation signals are produced during accident conditions.

Containment Phase "A" isolation occurs upon receipt of a safety injection

signal. The Phase "A" isolation signal isolates non-essential process

lines in order to minimize leakage of fission product radioactivity.

Containment Phase "B" isolation occurs upon receipt of a containment

pressure - High High signal and isolates the remaining process lines, except systems required for accident mitigation. In addition to the

isolation signals listed above, the purge and exhaust valves receive an

isolation signal on a containment high radiation condition. As a result, the

containment isolation valves (and blind flanges) help ensure that the

containment atmosphere will be isolated from the environment in the

event of a release of fission product radioactivity to the containment

atmosphere as a result of a Design Basis Accident (DBA).

The OPERABILITY requirements for containment isolation valves help

ensure that containment is isolated within the time limits assumed in the

safety analyses. Therefore, the OPERABILITY requirements provide

assurance that the containment function assumed in the safety analyses

will be maintained.

Containment Isolation Valves B 3.6.3 BASES (continued)

Watts Bar - Unit 2 B 3.6-14 (developmental)

A BACKGROUND (continued)

Reactor Building Purge Ventilation System The Reactor Building Purge Ventilation system operates to supply outside

air into the containment for ventilation and cooling or heating, to equalize

internal and external pressures and to reduce the concentration of noble

gases within containment prior to and during personnel access. The

supply and exhaust lines each contain two isolation valves. Because of

their large size and their exposure to higher containment pressure during

accident conditions, the 24 inch containment lower compartment purge

isolation valves are physically restricted to 50 degrees open.

Since the valves used in the Reactor Building Purge Ventilation System

are designed to meet the requirements for automatic containment

isolation valves, these valves may be opened as needed in MODES 1, 2, 3 and 4.

APPLICABLE

SAFETY ANALYSES The containment isolation valve LCO was derived from the 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. Therefore, the safety

analyses of any event requiring isolation of containment is applicable to

this LCO.

The DBAs that result in a significant release of radioactive material within

containment are a loss of coolant accident (LOCA) and a rod ejection

accident (Ref. 1). In the analyses 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 DBA analysis assumes that, within 60 seconds after the accident, isolation of the containment is complete and leakage terminated except

for the design leakage rate (L a) and for valves in the Essential Raw Cooling Water (ERCW) system and Co mponent Cooling System (CSS).

These valves are in liquid cont aining systems and have been evaluated

to have no impact on the DBA analysis. 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.

Containment Isolation Valves B 3.6.3 BASES (continued)

Watts Bar - Unit 2 B 3.6-15 (developmental)

A APPLICABLE SAFETY ANALYSES (continued)

The single failure criterion required to be imposed in the conduct of plant

safety analyses was considered in the original design of the containment

purge valves. Two valves in series on each purge line provide assurance

that both the supply and exhaust lines could be isolated even if a single

failure occurred. The inboard and outboard isolation valves on each line

are provided with redundant control and power trains, pneumatically

operated to open, and spring-loaded to close upon power loss or air

failure. This arrangement was designed to preclude common mode

failures from disabling both valves on a purge line.

The containment isolation valves satisfy Criterion 3 of the NRC Policy

Statement.

LCO Containment isolation valves form a part of the containment boundary.

The containment isolation valves' safety function is related to minimizing

the loss of reactor coolant inventory and establishing the containment

boundary during a DBA.

The automatic power operated isolation valves are required to have

isolation times within limits and to actuate on an automatic isolation

signal. The 24 inch containment lower compartment purge valves must

have blocks installed to prevent full opening. Blocked purge valves also

actuate on an automatic signal. The valves covered by this LCO are

listed along with their associated stroke times in the FSAR (Ref. 2).

The normally closed containment 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/devices are those listed in Reference 2.

Purge valves with resilient seals and shield building bypass valves meet

additional leakage rate requirements. The other containment isolation

valve leakage rates are addressed by LCO 3.6.1, "Containment," as

Type C testing.

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 containment boundary during

accidents.

Containment Isolation Valves B 3.6.3 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.6-16 (developmental)

A APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment. 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, 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 3.9.4, "Containment Penetrations."

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 (licensed or unlicensed) at the valve controls, who is in continuous

communication with the control room. In this way, the penetration can be

rapidly isolated when a need for containment isolation is indicated. For

valve controls located in the control room, an operator (other than the

Shift Operations Supervisor (SOS), ASOS, or the Operator at the

Controls) may monitor containment isolation signal status rather than be

stationed at the valve controls. Other secondary responsibilities which do

not prevent adequate monitoring of containment isolation signal status

may be performed by the operator provided his/her primary responsibility

is rapid isolation of the penetration when needed for containment

isolation. Use of the Unit Control Room Operator (CRO) to perform this

function should be limited to those situations where no other operator is

available.

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 third Note, which ensures

appropriate remedial actions are taken, if necessary, if the affected

systems are rendered inoperable by an inoperable containment isolation

valve.

In the event the isolation valve leakage results in exceeding the overall

containment leakage rate, Note 4 directs entry into the applicable

Conditions and Required Actions of LCO 3.6.1.

Containment Isolation Valves B 3.6.3 BASES (continued)

Watts Bar - Unit 2 B 3.6-17 (developmental)

B ACTIONS (continued)

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

flow paths is inoperable except for purge valve or shield building bypass

leakage not within limit, 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 a penetration flow path isolated in accordance with

Required Action A.1, the device used to isolate the penetration should be

the closest available one to containment. Required Action A.1 must be

completed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time is reasonable, considering the time required to isolate the penetration and the relative

importance of supporting containment OPERABILITY during MODES 1, 2, 3, and 4.

For affected penetration flow paths that cannot be restored to

OPERABLE status within the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 involves verification that those isolation devices outside containment and capable of being mispositioned are in the correct

position. The Completion Time of "Once per 31 days for isolation devices

outside containment" 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 containment, 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. For penetration flow paths with only one containment

isolation valve and a closed system, Condition C provides the appropriate

actions.

Containment Isolation Valves B 3.6.3 BASES (continued)

Watts Bar - Unit 2 B 3.6-18 (developmental)

B ACTIONS A.1 and A.2 (continued)

Required Action A.2 is modified by two Notes. Note 1 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. Note 2 applies to isolation devices that are locked, sealed, or otherwise secured in position and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since the function of locking, sealing, or securing components is to ensure that these devices are not inadvertently repositioned. 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 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . 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 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is consistent with the ACTIONS of

LCO 3.6.1. In the event the affected penetration is isolated in accordance

with Required 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

containment 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

control 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.

Containment Isolation Valves B 3.6.3 BASES (continued)

Watts Bar - Unit 2 B 3.6-19 (developmental)

B ACTIONS (continued)

C.1 and C.2 With one or more penetration flow paths with one containment isolation

valve inoperable, the inoperable valve flow path 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 flow path.

Required Action C.1 must be completed within the 4-hour 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 maintaining

containment integrity during MODES 1, 2, 3, and 4. In the event the

affected penetration flow path 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 periodic verification is necessary to

assure leak tightness of containment 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 because 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 only one containment

isolation valve and a closed system. This Note is necessary since this

Condition is written to specifically address those penetration flow paths in

a closed system. Required Action C.2 is modified by two Notes. Note 1 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 adm inistrative means is considered acceptable, since access to these areas is typically restricted. Note 2 applies to isolation devices that are locked, sealed, or otherwise secured in position and allows these devices to be verified closed by use of administrative means. Allowing verifi cation by administrative means is considered acceptable, since the function of locking, sealing, or securing components is to ensure that these devices are not inadvertently repositioned. Therefore, the probability of misalignment of these valves, once they have been verified to be in the proper position, is small.

Containment Isolation Valves B 3.6.3 BASES (continued)

Watts Bar - Unit 2 B 3.6-20 (developmental)

B ACTIONS D.1 With the shield building bypass leakage rate not within limit, the

assumptions of the safety analyses are not met. Therefore, the leakage

must be restored to within limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Restoration can be

accomplished by isolating the penetration(s) that caused the limit to be

exceeded by use of one closed and de-activated automatic valve, closed

manual valve, or blind flange. When a penetration is isolated the leakage

rate for the isolated penetration is assumed to be the actual pathway

leakage through the isolation device. If two isolation devices are used to

isolate the penetration, the leakage rate is assumed to be the lesser

actual pathway leakage of the two devices. The 4-hour Completion Time

is reasonable considering the time required to restore the leakage by

isolating the penetration(s) and the relative importance of shield building

bypass leakage to the overall containment function.

E.1, E.2, and E.3

In the event one or more containment purge valves in one or more

penetration flow paths are not within the purge valve leakage limits, purge

valve leakage must be restored to within limits, or the affected penetration

flow path must be isolated. The method of isolation must be by 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, closed manual valve, or blind flange. A

purge valve with resilient seals utilized to satisfy Required Action E.1

must have been demonstrated to meet the leakage requirements of

SR 3.6.3.5. The specified Completion Time is reasonable, considering

that one containment purge valve remains closed so that a gross breach

of containment does not exist.

In accordance with Required Action E.2, this penetration flow path must

be verified to be isolated on a periodic basis. The periodic verification is

necessary to ensure that containment penetrations required to be isolated

following an accident, which are 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 valve manipulation.

Rather, it involves verification that those isolation devices outside containment potentially capable of being mispositioned are in the correct

position. For the isolation devices inside containment, the time period

specified as "Prior to entering MODE 4 from MODE 5 if not performed

within the previous Containment Isolation Valves B 3.6.3 BASES (continued)

Watts Bar - Unit 2 B 3.6-21 (developmental)

B ACTIONS E.1, E.2, and E.3 (continued) 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.

For the containment purge valve with resilient seal that is isolated in

accordance with Required Action E.1, SR 3.6.3.5 must be performed at

least once every 92 days. This assures that degradation of the resilient

seal is detected and confirms that the leakage rate of the containment

purge valve does not increase during the time the penetration is isolated.

The normal Frequency for SR 3.6.3.5, 184 days, is based on an NRC

initiative, Generic Issue B-20 (Ref. 3). Since more reliance is placed on a

single valve while in this Condition, it is prudent to perform the SR more

often. Therefore, a Frequency of once per 92 days was chosen and has

been shown to be acceptable based on operating experience.

Required Action E.2 is modified by two Notes. Note 1 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. Note 2 applies to isolation devices that are locked, sealed, or otherwise secured in position and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since the function of locking, sealing, or securing components is to ensure that these devices are not inadvertently repositioned.

F.1 and F.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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . 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.

Containment Isolation Valves B 3.6.3 BASES (continued)

Watts Bar - Unit 2 B 3.6-22 (developmental)

B SURVEILLANCE REQUIREMENTS SR 3.6.3.1

This SR ensures that the purge valves are closed as required or, if open, open for an allowable reason. If a purge 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 purge valves are open for the reasons stated. The

valves may be opened for pressure control, ALARA or air quality

considerations for personnel entry, or for Surveillances that require the

valves to be open. All purge 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 containment, the containment

annulus, and the Main Steam Valve Vault Rooms, 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 of the containment boundary is within

design limits. This SR does not require any testing or valve manipulation.

Rather, it involves verification that those containment isolation valves in areas where the valves are capable of being mispositioned are in the

correct position. Since verification of valve position for these valves is

relatively easy, the 31 day Frequency is based on engineering judgment

and was chosen to provide added assurance of the correct positions.

The SR specifies that 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 verifi cation by administrative means is considered acceptable, since access to these areas is typically restricted

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.

Containment Isolation Valves B 3.6.3 BASES (continued)

Watts Bar - Unit 2 B 3.6-23 (developmental)

B SURVEILLANCE REQUIREMENTS SR 3.6.3.3

This SR requires verification that each containment isolation manual

valve and blind flange located inside containment, the containment

annulus, and the Main Steam Valve Vault Rooms, 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 of the containment boundary is within

design limits. For these containment isolation valves, 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 (e.g., locked valve program)

and may be verified by administrative means, because the probability of

their misalignment is low. The SR specifies that containment isolation

valves that are open under administrative controls are not required to

meet the SR during the time 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 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 power operated and automatic

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 analyses.

The isolation time and Frequency of this SR are in accordance with the

Inservice Testing Program or 92 days.

Containment Isolation Valves B 3.6.3 BASES (continued)

Watts Bar - Unit 2 B 3.6-24 (developmental)

A SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.3.5

For containment purge valves with resilient seals, additional leakage rate

testing beyond the test requirements of 10 CFR 50, Appendix J, Option B (Ref. 4), is required to ensure OPERABILITY.

Operating experience has demonstrated that this type of seal has the

potential to degrade in a shorter time period than do other seal types.

Based on this observation and the importance of maintaining this

penetration leak tight (due to the direct path between containment and the

environment), a Frequency of 184 days was established as part of the

NRC resolution of Generic Issue B-20, "Containment Leakage Due to

Seal Deterioration" (Ref. 3).

Additionally, this SR must be performed within 92 days after opening the

valve. The 92-day Frequency was chosen recognizing that cycling the

valve could introduce additional seal degradation (beyond that occurring

to a valve that has not been opened). Thus, decreasing the interval (from

184 days) is a prudent measure after a valve has been opened.

SR 3.6.3.6 Automatic containment isolation valves close on a containment isolation

signal to prevent leakage of radioactive material from containment

following a DBA. This SR ensures that each automatic containment

isolation valve will actuate to its isolation position on a containment

isolation signal. This Surveillance is not required for valves that are

locked, sealed, or otherwise secured in the required position under

administrative control. The 18-month Frequency is based on the need to

perform this Surveillance under the conditions that apply during a plant

outage and the potential for an unplanned transient if the Surveillance

were performed with the reactor at power.

Operating experience has shown that these components usually pass this

Surveillance when performed at the 18-month Frequency. Therefore, the

Frequency was concluded to be acceptable from a reliability standpoint.

Containment Isolation Valves B 3.6.3 BASES Watts Bar - Unit 2 B 3.6-25 (developmental)

A SURVEILLANCE REQUIREMENTS SR 3.6.3.7 Verifying that each 24 inch containment lower compartment purge valve

is blocked to restrict opening to 50 is required to ensure that the valves can close under DBA conditions within the times assumed in the analyses

of References 1 and 2. If a LOCA occurs, the purge valves must close to

maintain containment leakage within the values assumed in the accident

analysis. At other times when purge valves are required to be capable of

closing (e.g., during movement of irradiated fuel assemblies),

pressurization concerns are not present, thus the purge valves can be

fully open. The 18-month Frequency is appropriate because the blocking

devices are typically removed only during a refueling outage.

SR 3.6.3.8

This SR ensures that the combined leakage rate of all Shield Building

bypass leakage paths is less than or equal to the specified leakage rate.

This provides assurance that the assumptions in the safety analysis are

met. The as-left bypass leakage rate prior to the first startup after

performing a leakage test, requires calculation using maximum pathway

leakage (leakage through the worse of the two isolation valves). If the

penetration is isolated by use of one closed and de-activated automatic

valve, closed manual valve, or blind flange, then the leakage rate of the

isolated bypass leakage path is assumed to be the actual pathway

leakage through the isolation device. If both isolation valves in the

penetration are closed, the actual leakage rate is the lesser leakage rate

of the two valves. At all other times, the leakage rate will be calculated

using minimum pathway leakage.

The frequency is required by the Containment Leakage Rate Testing

Program. This SR simply imposes additional acceptance criteria.

Although not a part of L a , the Shield Building Bypass leakage path combined leakage rate is determined using the 10 CFR 50, Appendix J, Option B, Type B and C leakage rates for the applicable barriers.

Containment Isolation Valves B 3.6.3 BASES Watts Bar - Unit 2 B 3.6-26 (developmental)

A REFERENCES 1. Watts Bar FSAR, Section 15.0, "Accident Analysis." 2. Watts Bar FSAR, Section 6.2.4.2, "Containment Isolation System Design," and Table 6.2.4-1, "Containment Penetrations and Barriers." 3. Generic Issue B-20, "Containment Leakage Due to Seal Deterioration." 4. Title 10, Code of Federal Regulations, Part 50 Appendix J, Option B, "Primary Reactor Containment Leakage Testing for

Water-Cooled Power Reactors - Performance - Based

Requirements."

Containment Pressure B 3.6.4 (continued)

Watts Bar - Unit 2 B 3.6-27 (developmental)

B B 3.6 CONTAINMENT SYSTEMS B 3.6.4 Containment Pressure

BASES BACKGROUND The containment pressure is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a loss of

coolant accident (LOCA) or steam line break (SLB). These limits also prevent the containment pressure from exceeding the containment design

negative pressure differential (-2.0 psid) with respect to the shield building

annulus atmosphere in the event of inadvertent actuation of the

Containment Spray System or Air Return Fans.

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

outside these limits coincident with a Design Basis Accident (DBA), post

accident containment pressures could exceed calculated values.

APPLICABLE

SAFETY ANALYSES Containment internal pressure is an initial condition used in the DBA

analyses to establish the maximum peak containment internal pressure.

The limiting DBAs considered, relative to containment pressure, are the

LOCA and SLB, which are analyzed using computer pressure transients.

The worst case LOCA generates larger mass and energy release than

the worst case SLB. Thus, the LOCA event bounds the SLB event from

the containment peak pressure standpoint (Ref. 1).

The initial pressure condition used in the containment analysis was

15.0 psia. This resulted in a maximum peak pressure from a LOCA of

10.23 psig. The containment analysis (Ref. 1) shows that the maximum allowable internal containment pressure, P a (15.0 psig), bounds the calculated results from the limiting LOCA. The maximum containment

pressure resulting from the worst case LOCA, does not exceed the

containment design pressure, 13.5 psig.

Containment Pressure B 3.6.4 BASES (continued)

Watts Bar - Unit 2 B 3.6-28 (developmental)

A APPLICABLE SAFETY ANALYSES (continued)

The containment was also designed for an external pressure load

equivalent to 2.0 psig. The inadvertent actuation of the Containment

Spray System was analyzed to determine the resulting reduction in

containment pressure. The initial pressure condition used in this analysis

was -0.1 psig. This resulted in a minimum pressure inside containment of

1.4 psig, which is less than the design load.

For certain aspects of transient accident analyses, maximizing the

calculated containment pressure is not conservative. In particular, the

cooling effectiveness of the Emergency Core Cooling System during the

core reflood phase of a LOCA analysis increases with increasing

containment backpressure. Therefore, for the reflood phase, the

containment backpressure is calculated in a manner designed to

conservatively minimize, rather than maximize, the containment pressure

response in accordance with 10 CFR 50, Appendix K (Ref. 2).

Containment pressure satisfies Criterion 2 of the NRC Policy Statement.

LCO Maintaining containment pressure at 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 at greater than or equal to

the LCO lower pressure limit ensures that the containment will not exceed

the design negative differential pressure following the inadvertent

actuation of the Containment Spray System or Air Return Fans.

APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment. Since maintaining containment pressure within

limits is essential to ensure initial conditions assumed in the accident

analyses 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.

Containment Pressure B 3.6.4 BASES (continued)

Watts Bar - Unit 2 B 3.6-29 (developmental)

B ACTIONS A.1 When containment pressure is not within the limits of the LCO, it must be

restored to within these limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The Required Action is

necessary to return operation to within the bounds of the containment

analysis. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is consistent with the ACTIONS of

LCO 3.6.1, "Containment," which requires that containment be restored to

OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . 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

REQUIREMENTS SR 3.6.4.1

Verifying that containment pressure is within limits ( -0.1 and +0.3 psid relative to the annulus, value does not account for instrument error) ensures that plant operation remains within the limits assumed in the

containment analysis. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency of this SR was developed

based on operating experience related to trending of containment

pressure variations during the applicable MODES. Furthermore, the 12

hour Frequency is considered adequate in view of other indications

available in the control room, including alarms, to alert the operator to an

abnormal containment pressure condition.

REFERENCES 1. Watts Bar FSAR, Section 6.2.1, "Containment Functional Design." 2. Title 10, Code of Federal Regulations, Part 50, Appendix K, "ECCS Evaluation Models."

Containment Air Temperature B 3.6.5 (continued)

Watts Bar - Unit 2 B 3.6-30 (developmental)

A B 3.6 CONTAINMENT SYSTEMS B 3.6.5 Containment Air Temperature

BASES BACKGROUND The containment structure serves to contain radioactive material that may be released from the reactor core following a Design Basis Accident (DBA). The containment average air temperature is limited, during

normal operation, to preserve the initial conditions assumed in the

accident analyses for a loss of coolant accident (LOCA) or steam line

break (SLB).

The containment average air temperature limit is derived from the input

conditions used in the containment functional analyses and the

containment structure external pressure analyses. This LCO ensures that

initial conditions assumed in the analysis of containment response to a

DBA are not violated during plant operations. The total amount of energy

removed from containment by the Containment Spray and Cooling

systems during post accident conditions is dependent upon the energy

released to the containment due to the event, as well as the initial

containment temperature and pressure.

APPLICABLE

SAFETY ANALYSES Containment average air temperature is an initial condition 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 analyses for

containment (Ref. 1).

The limiting DBAs considered relative to containment OPERABILITY are

the LOCA and SLB. The DBA LOCA and SLB are analyzed using

computer codes designed to predict the resultant containment pressure

transients. No two DBAs are assumed to occur simultaneously or

consecutively. The postulated DBAs are analyzed with regard to

Engineered Safety Feature (ESF) systems, assuming the loss of one ESF

bus, which is the worst case single active failure, resulting in one train

each of Containment Spray System, Residual Heat Removal System, and

Air Return System being rendered inoperable.

Containment Air Temperature B 3.6.5 BASES (continued)

Watts Bar - Unit 2 B 3.6-31 (developmental)

B APPLICABLE SAFETY ANALYSES (continued)

The limiting DBA for the maximum peak containment air temperature is

an SLB. For the upper compartment, the initial containment average air

temperature assumed in this design basis analyses (Ref. 2) is 85 F. For the lower compartment, the initial average containment air temperature

assumed in this design basis analyses is 120 F. These temperatures result in a maximum containment air temperature.

The higher temperature limits are also considered in the depressurization

analyses to ensure that the minimum pressure limit is maintained

following an inadvertent actuation of the Containment Spray System for

both containment compartments.

The containment pressure transient is sensitive to the initial air mass in

containment and, therefore, to the initial containment air temperature.

The limiting DBA for establishing the maximum peak containment internal

pressure is a LOCA. The lower temperature limits, 85 F for the upper compartment and 100 F for the lower compartment, are used in this analysis to ensure that, in the event of an accident, the maximum

containment internal pressure will not be exceeded in either containment

compartment.

Containment average air temperature satisfies Criterion 2 of the NRC

Policy Statement.

LCO During a DBA, with an initial containment average air temperature within the LCO temperature limits, the resultant peak accident temperature is

maintained below the containment design temperature. As a result, the

ability of containment to perform its design function is ensured. In

MODES 2, 3 and 4, containment air temperature may be as low as 60 F (value does not account for instrument error) because the resultant

calculated peak containment accident pressure would not exceed the

design pressure due to a lesser amount of energy released from the pipe

break in these MODES (Ref. 3).

APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment. 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 average air temperature within the limit is not required in

MODE 5 or 6.

Containment Air Temperature B 3.6.5 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.6-32 (developmental)

B ACTIONS A.1 When containment average air temperature in the upper or lower

compartment is not within the limit of the LCO, the average air

temperature in the affected compartment must be restored to within limits

within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . This Required Action is necessary to return operation to

within the bounds of the containment analysis. The 8-hour Completion

Time is acceptable considering the sens itivity 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 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5

within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . 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

REQUIREMENTS SR 3.6.5.1 and SR 3.6.5.2

LCO 3.6.5 specifies that the containment average air temperature shall

be the following values which do not account for instrument error:

a. 85 F and 110 F for the containment upper compartment, and

b. 100 F and 120 F for the containment lower compartment.

Verifying that containment average air temperature is within the LCO

limits ensures that containment operation remains within the limits

assumed for the containment analyses. In order to determine the

containment average air temperature, a weighted average is calculated

using measurements taken at locations within the containment selected to

provide a representative sample of the overall containment atmosphere.

The 24-hour Frequency of these SRs is considered acceptable based on

observed slow rates of temperature increase within containment as a

result of environmental heat sources (due to the large volume of

containment). Furthermore, the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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.

Containment Air Temperature B 3.6.5 BASES Watts Bar - Unit 2 B 3.6-33 (developmental)

B REFERENCES 1. Watts Bar FSAR, Section 6.2, "Containment Systems." 2. Watts Bar System Description N3-30RB-4002R5, "Reactor Building Ventilation System." 3. Westinghouse Letter WAT-D-10698, dated November 23, 1999.

Containment Spray System B 3.6.6 (continued)

Watts Bar - Unit 2 B 3.6-34 (developmental)

A B 3.6 CONTAINMENT SYSTEMS B 3.6.6 Containment Spray System

BASES BACKGROUND The Containment Spray System provides containment atmosphere cooling to limit post accident pressure and temperature in containment to

less than the design values. Reduction of containment pressure helps

reduce the release of fission product radioactivity from containment to the

environment, in the event of a Design Basis Accident (DBA). The

Containment Spray System is designed to meet the requirements of

10 CFR 50, Appendix A, GDC 38, "Containment Heat Removal," GDC 39, "Inspection of Containment Heat Removal Systems," and GDC 40, "Testing of Containment Heat Removal Systems," (Ref. 1), or other

documents that were appropriate at the time of licensing (identified on a

plant specific basis).

The Containment Spray System consists of two separate trains of equal

capacity, each capable of meeting the system design basis spray

coverage. Each train includes a containment spray pump, one

containment spray heat exchanger, a spray header, nozzles, valves, and

piping. Each train is powered from a separate Engineered Safety Feature (ESF) bus. The refueling water storage tank (RWST) supplies borated

water to the Containment Spray System during the injection phase of

operation. In the recirculation mode of operation, containment spray

pump suction is transferred from the RWST to the containment

recirculation sump(s).

The diversion of a portion of the recirculation flow from each train of the

Residual Heat Removal (RHR) System to additional redundant spray

headers completes the Containment Spray System heat removal

capability. Each RHR train is capable of supplying spray coverage, if

required, to supplement the Containment Spray System.

The Containment Spray System and RHR System provide a spray of

subcooled borated water into the upper region of containment to limit the

containment pressure and temperature during a DBA. In the recirculation

mode of operation, heat is removed from the containment sump water by

the Containment Spray System and RHR heat exchangers. Each train of

the Containment Spray System, supplemented by a train of RHR spray, provides adequate spray coverage to meet the system design

requirements for containment heat removal.

Containment Spray System B 3.6.6 BASES (continued)

Watts Bar - Unit 2 B 3.6-35 (developmental)

A BACKGROUND (continued)

The Containment Spray System is ac tuated either automatically by a containment High-High pressure signal or manually. An automatic actuation starts the two containment spray pumps, opens the containment

spray pump discharge valves, and begins the injection phase. A manual

actuation of the Containment Spray System requires the operator to

actuate two separate switches on the main control board to begin the

same sequence. The injection phase continues until an RWST level

Low-Low alarm is received. The Low-Low alarm for the RWST signals

the operator to manually align the system to the recirculation mode. The

Containment Spray System in the 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.

The RHR spray operation is initiated manually, when required by the

emergency operating procedures, after the Emergency Core Cooling

System (ECCS) is operating in the recirculation mode. The RHR sprays

are available to supplement the Containment Spray System, if required, in

limiting containment pressure. This additional spray capacity would

typically be used after the ice bed has been depleted and in the event that

containment pressure rises above a pre-determined limit.

The Containment Spray System is an ESF system. It is designed to

ensure that the heat removal capability required during the post accident

period can be attained.

The operation of the ice condenser is adequate to assure pressure

suppression during the initial blowdown of steam and water from a DBA.

During the post blowdown period, the Air Return System (ARS) is

automatically started. The ARS returns upper compartment air through

the divider barrier to the lower compartment. This serves to equalize

pressures in containment and to continue circulating heated air and

steam through the ice condenser, where heat is removed by the

remaining ice and by the Containment Spray System after the ice has

melted.

The Containment Spray System limits the temperature and pressure that

could be expected following a DBA. Protection of containment integrity

limits leakage of fission product radioactivity from containment to the

environment.

Containment Spray System B 3.6.6 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.6-36 (developmental)

B APPLICABLE SAFETY ANALYSES The limiting DBAs considered relative to containment are the loss of

coolant accident (LOCA) and the steam line break (SLB). The DBA

LOCA and SLB are analyzed using computer codes designed to predict

the resultant containment pressure and temperature transients. No two

DBAs are assumed to occur simultaneously or consecutively. The

postulated DBAs are analyzed, in 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, the RHR

System, and the ARS being rendered inoperable (Ref. 2).

The DBA analyses show that the maximum peak containment pressure of

10.23 psig results from the LOCA analysis, and is calculated to be less than the containment design pressure. The maximum peak containment

atmosphere temperature results from the SLB analysis. The calculated

transient containment atmosphere temperatures are acceptable for the

DBA SLB.

The modeled Containment Spray System actuation from the containment analysis is based on a response time associated with exceeding the

containment High-High pressure signal setpoint to achieving full flow

through the containment spray nozzles. A delayed response time

initiation provides conservative analyses of peak calculated containment

temperature and pressure responses. The Containment Spray System

total response time of 234 seconds is composed of signal delay, diesel generator startup, and system startup time.

For certain aspects of transient accident analyses, maximizing the

calculated containment pressure is not conservative. In particular, the

ECCS cooling effectiveness during the core reflood phase of a LOCA

analysis increases with increasing containment backpressure. For these

calculations, the containment backpressure is calculated in a manner

designed to conservatively minimize, rather than maximize, the calculated

transient containment pressures in accordance with 10 CFR 50, Appendix K (Ref. 3).

Inadvertent actuation of the Containment Spray System is evaluated in

the analysis, and the resultant reduction in containment pressure is

calculated. The maximum calculated steady state pressure differential

relative to the Shield Building annulus is 1.4 psid, which is below the

containment design external pressure load of 2.0 psid.

The Containment Spray System satisfies Criterion 3 of the NRC Policy

Statement.

Containment Spray System B 3.6.6 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.6-37 (developmental)

B LCO During a DBA, one train of Containment Spray System and RHR Spray System is required to provide the heat removal capability assumed in the safety analyses. To ensure that these requirements are met, two

containment spray trains and two RHR spray trains must be OPERABLE

with power from two safety related, independent power supplies.

Therefore, in the event of an accident, at least one train in each system

operates.

Each containment spray train typically includes a spray pump, header, valves, a heat exchanger, nozzles, piping, instruments, and controls to

ensure an OPERABLE flow path capable of taking suction from the

RWST upon an ESF actuation signal and transferring suction to the containment sump. This suction path realignment is accomplished by manual operator action upon receipt of a Low-Low level alarm for the RWST.

Each RHR spray train includes a pump, header, valves, a heat

exchanger, nozzles, piping, instruments, and controls to ensure an

OPERABLE flow path capable of taking suction from the containment

sump and supplying flow to the spray header.

APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment and an increase in containment pressure and

temperature requiring the operation of the Containment Spray System.

A Note has been added which states the RHR spray trains are not

required in MODE 4. The containment spray system does not require

supplemental cooling from the RHR spray in MODE 4.

In MODES 5 and 6, the probability and consequences of these events are

reduced because of the pressure and temperature limitations of these

MODES. Thus, the Containment Spray System is not required to be

OPERABLE in MODE 5 or 6.

ACTIONS A.1 and B.1

With one containment spray train and/or RHR spray train inoperable, the

affected train must be restored to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The components in this degraded condition are capable of providing

100% of the heat removal needs after an accident. The 72-hour

Completion Time was developed taking into account the redundant heat

removal capabilities afforded by the OPERABLE train and the low

probability of a DBA occurring during this period.

Containment Spray System B 3.6.6 BASES (continued)

Watts Bar - Unit 2 B 3.6-38 (developmental)

B ACTIONS (continued)

C.1 and C.2 If the affected containment spray train and/or RHR 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months . 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. The extended interval to reach

MODE 5 allows additional time and is reasonable when considering that

the driving force for a release of radioactive material from the Reactor

Coolant System is reduced in MODE 3.

SURVEILLANCE

REQUIREMENTS SR 3.6.6.1

Verifying the correct alignment of manual, power operated, and automatic

valves, excluding check valves, in the Containment Spray System

provides assurance that the proper flow path exists for Containment

Spray System operation. This SR does not apply to valves that are

locked, sealed, or otherwise secured in position since they were verified

in the correct position prior to being secured. This SR does not require

any testing or valve manipulation. Rather, it involves verification that those valves outside containment and capable of potentially being mis-

positioned, are in the correct position.

SR 3.6.6.2

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 head are normal tests of centrifugal pump performance

required by the American Society of Mechanical Engineers (ASME) OM

Code (Ref. 4). Since the containment spray pumps cannot be tested with

flow through the spray headers, they are tested on bypass flow. This test

confirms one point on the pump design curve and is indicative of overall

performance. Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal

performance. The Frequency of this SR is in accordance with the

Inservice Testing Program.

Containment Spray System B 3.6.6 BASES (continued)

Watts Bar - Unit 2 B 3.6-39 (developmental)

A SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.6.3 and SR 3.6.6.4

These SRs require verification that each automatic containment spray

valve actuates to its correct position and each containment spray pump

starts upon receipt of an actual or simulated containment spray actuation

signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative

control. Containment spray pump start verification may be performed by

testing breaker actuation without pump start (breaker is racked out in its "test position") and observation of the local or remote pump start lights (breaker energization light). The 18-month Frequency is based on the

need to perform these Surveillances under the conditions that apply

during a plant outage and the potential for an unplanned transient if the

Surveillances were performed with the reactor at power. Operating

experience has shown these components usually pass the Surveillances

when performed at the 18-month Frequency. Therefore, the Frequency

was concluded to be acceptable from a reliability standpoint.

The surveillance of containment sump isolation valves is also required by

SR 3.6.6.3. A single surveillance may be used to satisfy both

requirements.

SR 3.6.6.5

With the containment spray inlet valves closed and the spray header

drained of any solution, low pressure air or smoke can be blown through

test connections. This SR ensures that each spray nozzle required by the

design bases is unobstructed and that spray coverage of the containment

during an accident is not degraded. Because of the passive design of the

nozzle, a test at the first refueling and at 10 year intervals are considered

adequate to detect obstruction of the spray nozzles.

SR 3.6.6.6

The Surveillance descriptions from Bases 3.5.2 for SR 3.5.2.2 and

SR 3.5.2.4 apply as applicable to the RHR spray system.

Containment Spray System B 3.6.6 BASES (continued)

Watts Bar - Unit 2 B 3.6-40 (developmental)

A REFERENCES 1. Title 10, Code of Federal Regulations, Part 50, Appendix A, "General Design Criterion (GDC) 38, "Containment Heat Removal," GDC 39, "Inspection of Containment Heat Removal System,"

GDC 40, "Testing of Containment Heat Removal Systems, and

GDC 50, "Containment Design Basis." 2. Watts Bar FSAR, Section 6.2, "Containment Systems." 3. Title 10, Code of Federal Regulations, Part 50, Appendix K, "ECCS Evaluation Models." 4. American Society of Mechanical Engineers (ASME) OM Code, "Code for Operation and Maintenance of Nuclear Power Plants."

Hydrogen Recombiners B 3.6.7 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.6-41 (developmental)

B B 3.6 CONTAINMENT SYSTEMS B 3.6.7 The Bases for Specification 3.6.7 have been Deleted

HMS B 3.6.8 (continued)

Watts Bar - Unit 2 B 3.6-42 (developmental)

A B 3.6 CONTAINMENT SYSTEMS B 3.6.8 Hydrogen Mitigation System (HMS)

BASES BACKGROUND The HMS consists of two groups of 34 ignitors distributed throughout the containment. The HMS reduces the potential for breach of primary

containment due to a hydrogen oxygen reaction in post accident

environments. The HMS is required by 10 CFR 50.44, "Standards for

Combustible Gas Control Systems in Light-Water-Cooled Reactors" (Ref. 1), and Appendix A, GDC 41, "Containment Atmosphere Cleanup" (Ref. 2), to reduce the hydrogen concentration in the primary containment

following a degraded core accident. The HMS must be capable of

handling an amount of hydrogen equivalent to that generated from a

metal water reaction involving 75% of the fuel cladding surrounding the active fuel region (excluding the plenum volume).

10 CFR 50.44 (Ref. 1) requires plants with ice condenser containments to

install suitable hydrogen control systems that would accommodate an

amount of hydrogen equivalent to that generated from the reaction of

75% of the fuel cladding with water. The HMS provides this required

capability. This requirement was placed on ice condenser plants because

of their small containment volume and low design pressure (compared

with pressurized water reactor dry containments). Calculations indicate

that if hydrogen equivalent to that generated from the reaction of 75% of

the fuel cladding with water were to collect in the primary containment, the resulting hydrogen concentration would be far above the lower

flammability limit such that, if ignited from a random ignition source, the

resulting hydrogen burn would seriously challenge the containment and

safety systems in the containment.

The HMS is based on the concept of controlled ignition using thermal

ignitors, designed to be capable of functioning in a post accident

environment, seismically supported, and capable of actuation from the

control room. A total of 68 ignitors are distributed throughout the various

regions of containment in which hydrogen could be released or to which it

could flow in significant quantities. The ignitors are arranged in two independent trains such that each containment region has at least two ignitors, one from each train, controlled and powered redundantly so

that ignition would occur in each region even if one train failed to

energize.

HMS B 3.6.8 BASES (continued)

Watts Bar - Unit 2 B 3.6-43 (developmental)

A BACKGROUND (continued)

When the HMS is initiated, the ignitor elements are energized and heat up to a surface temperature 1700 F. At this temperature, they ignite the hydrogen gas that is present in the airspace in the vicinity of the ignitor.

The HMS depends on the dispersed location of the ignitors so that local

pockets of hydrogen at increased concentrations would burn before

reaching a hydrogen concentration significantly higher than the lower

flammability limit. Hydrogen ignition in the vicinity of the ignitors is

assumed to occur when the local hydrogen concentration reaches a

minimum 5.0 volume percent (v/o).

APPLICABLE

SAFETY ANALYSES The HMS causes hydrogen in containment to burn in a controlled manner

as it accumulates following a degraded core accident (Ref. 3). Burning

occurs at the lower flammability concentration, where the resulting

temperatures and pressures are relatively benign. Without the system, hydrogen could build up to higher concentrations that could result in a

violent reaction if ignited by a random ignition source after such a buildup.

The hydrogen ignitors are not included for mitigation of a Design Basis

Accident (DBA) because an amount of hydrogen equivalent to that

generated from the reaction of 75% of the fuel cladding with water is far in excess of the hydrogen calculated for the limiting DBA loss of coolant accident (LOCA). The hydrogen concentration resulting from a DBA can

be maintained less than the flammability limit using the hydrogen

recombiners. The hydrogen ignitors, however, have been shown by

probabilistic risk analysis to be a significant contributor to limiting the

severity of accident sequences that are commonly found to dominate risk

for plants with ice condenser containments. As such, the hydrogen

ignitors are considered to be risk significant in accordance with the NRC

Policy Statement.

HMS B 3.6.8 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.6-44 (developmental)

A LCO Two HMS trains must be OPERABLE with power from two independent, safety related power supplies.

For this plant, an OPERABLE HMS train consists of 33 of 34 ignitors

energized on the train.

Operation with at least one HMS train ensures that the hydrogen in

containment can be burned in a controlled manner. Unavailability of both

HMS trains could lead to hydrogen buildup to higher concentrations, which could result in a violent reaction if ignited. The reaction could take

place fast enough to lead to high temperatures and overpressurization of

containment and, as a result, breach containment or cause containment

leakage rates above those assumed in the safety analyses. Damage to

safety related equipment located in containment could also occur.

APPLICABILITY Requiring OPERABILITY in MODES 1 and 2 for the HMS ensures its immediate availability after safety injection and scram actuated on a

LOCA initiation. In the post accident environment, the two HMS

subsystems are required to control the hydrogen concentration within

containment to near its flammability limit of 4.0 v/o assuming a worst case

single failure. This prevents overpressurization of containment and

damage to safety related equipment and instruments located within

containment.

In MODES 3 and 4, both the hydrogen production rate and the total

hydrogen production after a LOCA would be significantly less than that calculated for the DBA LOCA. Also, because of the limited time in these MODES, the probability of an accident requiring the HMS is low.

Therefore, the HMS is not required in MODES 3 and 4.

In MODES 5 and 6, the probability and consequences of a LOCA are

reduced due to the pressure and temperature limitations of these

MODES. Therefore, the HMS is not required to be OPERABLE in

MODES 5 and 6.

HMS B 3.6.8 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.6-45 (developmental)

A ACTIONS A.1 and A.2 With one HMS train inoperable, the inoperable train must be restored to

OPERABLE status within 7 days or the OPERABLE train must be verified

OPERABLE frequently by performance of SR 3.6.8.1. The 7-day

Completion Time is based on the low probability of the occurrence of a

degraded core event that would generate hydrogen in amounts equivalent

to a metal water reaction of 75% of the core cladding, the length of time

after the event that operator action would be required to prevent hydrogen

accumulation from exceeding this limit, and the low probability of failure of

the OPERABLE HMS train. Alternative Required Action A.2, by frequent surveillances, provides assurance that the OPERABLE train continues to be OPERABLE.

B.1 Condition B is one containment region with no OPERABLE hydrogen

ignitor. Thus, while in Condition B, or in Conditions A and B

simultaneously, there would always be ignition capability in the adjacent

containment regions that would provide redundant capability by flame

propagation to the region with no OPERABLE ignitors.

Required Action B.1 calls for the restoration of one hydrogen ignitor in

each region to OPERABLE status within 7 days. The 7-day Completion

Time is based on the same reasons given under Required Action A.1.

C.1 If the HMS subsystem(s) 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The Completion Time of

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach MODE 3

from full power conditions in an orderly manner and without challenging

plant systems.

HMS B 3.6.8 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.6-46 (developmental)

A SURVEILLANCE REQUIREMENTS SR 3.6.8.1

This SR confirms that 33 of 34 hydrogen ignitors can be successfully energized in each train. The ignitors are simple resistance elements.

Therefore, energizing provides assurance of OPERABILITY. The

allowance of one inoperable hydrogen ignitor is acceptable because, although one inoperable hydrogen ignitor in a region would compromise

redundancy in that region, the containment regions are interconnected so

that ignition in one region would cause burning to progress to the others (i.e., there is overlap in each hydrogen ignitor's effectiveness between

regions). The Frequency of 92 days has been shown to be acceptable through operating experience.

SR 3.6.8.2

This SR confirms that the two inoperable hydrogen ignitors allowed by

SR 3.6.8.1 (i.e., one in each train) are not in the same containment

region. The containment regions and hydrogen ignitor locations are

provided in Reference 3. The Frequency of 92 days is acceptable based

on the Frequency of SR 3.6.8.1, which provides the information for

performing this SR.

SR 3.6.8.3

A more detailed functional test is performed every 18 months to verify

system OPERABILITY. Each glow pl ug is visually examined to ensure that it is clean and that the electrical circuitry is energized. All ignitors (glow plugs), including normally inaccessible ignitors, are visually

checked for a glow to verify that they are energized. Additionally, the

surface temperature of each glow plug is measured to be 1700 F to demonstrate that a temperature sufficient for ignition is achieved. The 18-

month Frequency is based on the need to perform this Surveillance under

the conditions that apply during a plant outage and the potential for an

unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has s hown that these components usually pass the SR when performed at the 18-month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be

acceptable from a reliability standpoint.

HMS B 3.6.8 BASES (continued)

Watts Bar - Unit 2 B 3.6-47 (developmental)

A REFERENCES

1. Title 10, Code of Federal Regulations, Part 50.44, "Standards for Combustible Gas Control Systems in Light Water-Cooled Power Reactors." 2. Title 10, Code of Federal Regulations, Part 50, Appendix A, General Design Criterion 41, "Containment Atmosphere Cleanup." 3. Watts Bar FSAR, Section 6.2.5A, "Hydrogen Mitigation System Description."

EGTS B 3.6.9 (continued)

Watts Bar - Unit 2 B 3.6-48 (developmental)

A B 3.6 CONTAINMENT SYSTEMS

B 3.6.9 Emergency Gas Treatment System (EGTS)

BASES BACKGROUND The EGTS is required by 10 CFR 50, Appendix A, GDC 41, "Containment Atmosphere Cleanup" (Ref. 1), to ensure that radioactive materials that leak from the primary containment into the shield building (secondary containment) following a Design Basis Accident (DBA) are filtered and adsorbed prior to exhausting to the environment.

The containment has a secondary containment called the shield building, which is a concrete structure that surrounds the steel primary containment vessel. Between the containment vessel and the shield building inner wall is an annular space that collects any containment leakage that may occur following a loss of coolant accident (LOCA). This space also allows for periodic inspection of the outer surface of the steel containment vessel.

The EGTS establishes a negative pressure in the annulus between the shield building and the steel containment vessel. Filters in the system then control the release of radioactive contaminants to the environment. Shield building OPERABILITY is required to ensure retention of primary containment leakage and proper operation of the EGTS.

The EGTS consists of two separate and redundant trains. Each train includes a heater, a prefilter, moisture separators, a high efficiency particulate air (HEPA) filter, an activated charcoal adsorber section for removal of radioiodines, and a fan. Ductwork, valves and/or dampers, and instrumentation also form part of the system. The moisture separators function to reduce the moisture content of the airstream.

A second bank of HEPA filters follows the adsorber section to collect carbon fines and provide backup in case of failure of the main HEPA filter bank. Only the upstream HEPA filter and the charcoal adsorber section are credited in the analysis. The system initiates and maintains a negative air pressure in the shield building by means of filtered exhaust ventilation of the shield building following receipt of a safety injection (SI) signal. The system is described in Reference 2.

EGTS B 3.6.9BASES (continued)

Watts Bar - Unit 2 B 3.6-49 (developmental)

B BACKGROUND (continued) The prefilters remove large particles in the air, and the moisture separators remove entrained water droplets present, to prevent excessive loading of the HEPA filters and charcoal absorbers. Heaters are included to reduce the relative humidity of the airstream on systems that operate in high humidity. Continuous operation of each train, for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months

per month, with heaters on, reduces moisture buildup on their HEPA filters and adsorbers. Cross-over flow ducts are provided between the two trains to allow the active train to draw air through the inactive train and cool the air to keep the charcoal beds on the inactive train from becoming too hot due to absorption of fission products.

The containment annulus vacuum fans maintain the annulus at -5 inches water gauge vacuum during normal operations. During accident conditions, the containment annulus vacuum fans are isolated from the air cleanup portion of the system.

The EGTS reduces the radioactive content in the shield building atmosphere following a DBA. Loss of the EGTS could cause site boundary doses, in the event of a DBA, to exceed the values given in the licensing basis.

APPLICABLE

SAFETY ANALYSES The EGTS design basis is established by the consequences of the limiting DBA, which is a LOCA. The accident analysis (Ref. 3) considers two different single failure scenarios. The first one assumes that only one train of the EGTS is functional due to a postulated single failure that disables the other train. An alternate scenario assumes a single failure of the pressure control loop associated with one train of PCOs. The first scenario is bounding for thyroid dose while the alternate scenario is bounding for beta and gamma doses. The accident analysis accounts for the reduction in airborne radioactive material provided by the number of filter trains in operation for each failure scenario. The amount of fission products available for release from containment is determined for a LOCA.

The safety analysis conservatively assumes the annulus is at atmospheric pressure prior to the LOCA. The analysis further assumes that upon receipt of a Containment Isolation Phase A (CIA) signal from the RPS, the EGTS fans automatically start and achieve a minimum flow of 3600 cfm (per train) within 18 seconds (20 seconds from the initiating event.) This does not include 10 seconds for diesel generator startup.

EGTS B 3.6.9BASES (continued)Watts Bar - Unit 2 B 3.6-50 (developmental)

B APPLICABLE SAFETY ANALYSES (continued) The analysis shows that the annulus pressure will rise to a positive value and then decrease to the EGTS control point for a single failure of one EGTS train, or slightly more negative for a single failure of a pressure control loop associated with one train of PCOs. The normal alignment for both EGTS control loops is the A-Auto position. With both EGTS control loops in A-Auto, both trains will function upon initiation of a CIA signal. In the event of a LOCA, the annulus va cuum control system isolates and both trains of the EGTS pressure control loops will be placed in service to maintain the required negative pressure. If annulus vacuum is lost during normal operations, the A-Auto position is unaffected by the loss of vacuum. This operational configuration is acceptable because the accident dose analysis conservatively assumes the annulus is at atmospheric pressure at event initiation.

The EGTS satisfies Criterion 3 of the NRC Policy Statement.

LCO In the event of a DBA, one EGTS train is required to provide the minimum particulate iodine removal assumed in the safety analysis. Two trains of the EGTS must be OPERABLE to ensure that at least one train will operate, assuming that the other train is disabled by a single active failure.

See TS Bases 3.6.15, "Shield Building," for additional information on EGTS.

APPLICABILITY In MODES 1, 2, 3, and 4, a D BA could lead to fission product release to containment that leaks to the shield building. The large break LOCA, on which this system's design is based, is a full power event. Less severe LOCAs and leakage still require the system to be OPERABLE throughout these MODES. The probability and severity of a LOCA decrease as core power and Reactor Coolant System pressure decrease. With the reactor shut down, the probability of release of radioactivity resulting from such an accident is low.

In MODES 5 and 6, the probability and consequences of a DBA are low due to the pressure and temperature limitations in these MODES. Under these conditions, the Filtration System is not required to be OPERABLE (although one or more trains may be operating for other reasons, such as habitability during maintenance in the shield building annulus).

EGTS B 3.6.9BASES (continued)Watts Bar - Unit 2 B 3.6-51 (developmental)

A ACTIONS (continued)

A.1 With one EGTS train inoperable, the inoperable train must be restored to OPERABLE status within 7 days. The components in this degraded condition are capable of providing 100% of the iodine removal needs after a DBA. The 7-day Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant EGTS train and the low probability of a DBA occurring during this period. The Completion Time is adequate to make most repairs.

B.1 and B.2

If the EGTS 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . 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 REQUIREMENTS SR 3.6.9.1 Operating each EGTS train for 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months 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 for corrective action. Operation with the heaters on (automatic heater cycling to maintain temperature) for 10 continuous hours eliminates moisture on the adsorbers and HEPA filters. Experience from filter testing at operating units indicates that the 10-hour period is

adequate for moisture elimination on the adsorbers and HEPA filters.

The 31-day Frequency was developed in consideration of the known reliability of fan motors and controls, the two train redundancy available.

SR 3.6.9.2

This SR verifies that the required EGTS filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP - Technical Specification Section 5.7.2.14). The EGTS filter tests are in accordance with Regulatory Guide 1.52 (Ref. 4). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).

EGTS B 3.6.9BASES (continued)Watts Bar - Unit 2 B 3.6-52 (developmental)

B SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.9.2 (continued)

Specific test frequencies and additional information are discussed in detail in the VFTP. It should be noted that for the EGTS, the VFTP pressure drop value across the entire filtration unit does not account for

instrument error.

SR 3.6.9.3

The automatic startup ensures that each EGTS train responds properly.

This testing includes the automatic swapping logic of the EGTS pressure control isolation valves in response to the actuation signal. Performance of this swapping logic test will ensure the availability of EGTS functions in the event of an initial single failure of one of the pressure control loops. The 18-month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at the 18-month Frequency. Therefore the Frequency was concluded to be acceptable from a reliability standpoint. Furthermore, the SR interval was developed considering that the EGTS equipment OPERABILITY is demonstrated at a 31-day Frequency by SR 3.6.9.1.

SR 3.6.9.4

The proper functioning of the fans, dampers, filters, adsorbers, etc., as a system is verified by the ability of each train to produce the required system flow rate within the specified timeframe. The 18-month Frequency on a STAGGERED TEST BASIS is consistent with Regulatory Guide 1.52 (Ref. 4) guidance for functional testing.

EGTS B 3.6.9BASES (continued)Watts Bar - Unit 2 B 3.6-53 (developmental)

B REFERENCES 1. Title 10, Code of Federal Regulations, Part 50, Appendix A, General Design Criterion 41, "Containment Atmosphere Cleanup." 2. Watts Bar FSAR, Section 6.5, "Fission Product Removal and Control Systems." 3. Watts Bar FSAR, Section 15.0, "Accident Analysis." 4. Regulatory Guide 1.52, Rev. 2, "Design, Testing and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmospheric Cleanup System Air Filtration and Absorption Units of Light-Water Cooled Nuclear Power Plants."

ARS B 3.6.10 (continued)

B 3.6 CONTAINMENT SYSTEMS B 3.6.10 Air Return System (ARS)

BASES BACKGROUND The ARS is designed to assure the rapid return of air from the upper to the lower containment compartment after the initial blowdown following a Design Basis Accident (DBA). The return of this air to the lower

compartment and subsequent recirculation back up through the ice

condenser assists in cooling the containment atmosphere and limiting

post accident pressure and temperature in containment to less than

design values. Limiting pressure and temperature reduces the release of

fission product radioactivity from containment to the environment in the

event of a DBA.

The ARS provides post accident hydrogen mixing in selected areas of containment. The ARS draws air from the dome of the containment vessel, from the reactor cavity, and from the ten dead ended (pocketed)

spaces in the containment where there is potential for the accumulation

of hydrogen. The minimum design flow from each potential hydrogen

pocket is sufficient to limit the local concentration of hydrogen.

The ARS consists of two separate trains of equal capacity, each capable

of meeting the design bases. Each train includes a 100% capacity air

return fan, associated damper, and hydrogen collection headers. Each

train is powered from a separate Engineered Safety Features (ESF) bus.

The ARS fans are automatically started by the containment isolation

Phase B signal 8 to 10 minutes after the containment pressure reaches

the pressure setpoint. The time delay ensures that no energy released

during the initial phase of a DBA will bypass the ice bed through the ARS

fans into the upper containment compartment.

After starting, the fans displace air from the upper compartment to the

lower compartment, thereby returning the air that was displaced by the high energy line break blowdown from the lower compartment and equalizing pressures throughout containment. After discharge into the lower compartment, air flows with steam produced by residual heat through the ice condenser doors into the ice condenser compartment

where the steam portion of the flow is condensed. The air flow returns to

the upper compartment through the top deck doors in the upper portion

of the ice condenser compartment. The ARS fans operate continuously

after actuation, circulating air through the containment volume and Watts Bar - Unit 2 B 3.6-54 (developmental)

A ARS B 3.6.10 BASES (continued)

Watts Bar - Unit 2 B 3.6-55 (developmental)

A BACKGROUND (continued) purging all potential hydrogen pockets in containment. When the containment pressure falls below a predetermined value, the ARS fans

are manually de-energized. Thereafter, the fans are manually cycled on and off if necessary to control any additional containment pressure

transients.

The ARS also functions, after all the ice has melted, to circulate any

steam still entering the lower compartment to the upper compartment

where the Containment Spray System can cool it.

The ARS is an ESF system. It is designed to ensure that the heat

removal capability required during the post accident period can be

attained. The operation of the ARS, in conjunction with the ice bed, the

Containment Spray System, and the Residual Heat Removal (RHR)

System spray, provides the required heat removal capability to limit post accident conditions to less than the containment design values.

APPLICABLE

SAFETY ANALYSES The limiting DBAs considered relative to containment temperature and

pressure are the loss of coolant accident (LOCA) and the steam line

break (SLB). The LOCA and SLB are analyzed using computer codes

designed to predict the resultant containment pressure and temperature

transients. DBAs are assumed not to occur simultaneously or

consecutively. The postulated DBAs are analyzed, in regard to ESF

systems, assuming the loss of one ESF bus, which is the worst case

single active failure and results in one train each of the Containment

Spray System, RHR System, and ARS being inoperable (Ref. 1). The

DBA analyses show that the maximum peak containment pressure results

from the LOCA analysis and is calculated to be less than the containment design pressure.

For certain aspects of transient accident analyses, maximizing the

calculated containment pressure is not conservative. In particular, the

cooling effectiveness of the Emergency Core Cooling System during the

core reflood phase of a LOCA analysis increases with increasing

containment backpressure. For these calculations, the containment

backpressure is calculated in a manner designed to conservatively

minimize, rather than maximize, the calculated transient containment

pressures, in accordance with 10 CFR 50, Appendix K (Ref. 2).

ARS B 3.6.10 BASES (continued)

Watts Bar - Unit 2 B 3.6-56 (developmental)

A APPLICABLE SAFETY ANALYSES (continued)

The modeled ARS actuation from the containment analysis is based upon

a response time associated with exceeding the containment pressure

High-High signal setpoint to achieving full ARS air flow. A delayed

response time initiation provides conservative analyses of peak

calculated containment temperature and pressure responses. The ARS

total response time of 540

+ 60 seconds consists of the built in signal delay.

The ARS satisfies Criterion 3 of the NRC Policy Statement.

LCO In the event of a DBA, one train of the ARS is required to provide the minimum air recirculation for heat removal and hydrogen mixing assumed

in the safety analyses. To ensure this requirement is met, two trains of

the ARS must be OPERABLE. This will ensure that at least one train will

operate, assuming the worst case single failure occurs, which is in the

ESF power supply.

APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the operation of the ARS. Therefore, 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, the ARS is not required to be OPERABLE in these

MODES.

ACTIONS A.1 If one of the required trains of the ARS is inoperable, it must be restored

to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The components in this degraded

condition are capable of providing 100% of the flow capability after an

accident. The 72-hour Completion Time was developed taking into

account the redundant flow and hydrogen mixing capability of the

OPERABLE ARS train and the low probability of a DBA occurring in this

period.

ARS B 3.6.10 BASES (continued)

Watts Bar - Unit 2 B 3.6-57 (developmental)

A ACTIONS (continued)

B.1 and B.2 If the ARS 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . 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

REQUIREMENTS SR 3.6.10.1

Verifying that each ARS fan starts on an actual or simulated actuation

signal, after a delay of 8.0 minutes and 10.0 minutes, and operates for 15 minutes is sufficient to ensure that all fans are OPERABLE and that all associated controls and time delays are functioning properly. It also

ensures that blockage, fan and/or motor failure, or excessive vibration

can be detected for corrective action. The 92 day Frequency was

developed considering the known reliability of fan motors and controls

and the two train redundancy available.

SR 3.6.10.2

Verifying ARS fan motor current with the return air backdraft dampers

closed confirms one operating condition of the fan. This test is indicative

of overall fan motor performance. Such inservice tests confirm

component OPERABILITY, trend performance, and detect incipient

failures by indicating abnormal performance. The Frequency of 92 days

conforms with the testing requirements for similar ESF equipment and

considers the known reliability of fan motors and controls and the two

train redundancy available.

SR 3.6.10.3

Verifying the OPERABILITY of the air return damper to the proper

opening torque (Ref. 3) provides assurance that the proper flow path will

exist when the fan is started. By applying the correct torque to the

damper shaft, the damper operation can be confirmed. The Frequency of

92 days was developed considering the importance of the dampers, their

location, physical environment, and probability of failure. Operating

experience has also shown this Frequency to be acceptable.

ARS B 3.6.10 BASES Watts Bar - Unit 2 B 3.6-58 (developmental)

A REFERENCES

1. Watts Bar FSAR, Section 6.8, "Air Return Fans." 2. Title 10, Code of Federal Regulations, Part 50, Appendix K, "ECCS Evaluation Models." 3. System Description N3-30RB-4002.

Ice Bed B 3.6.11 (continued)

Watts Bar - Unit 2 B 3.6-59 (developmental)

A B 3.6 CONTAINMENT SYSTEMS B 3.6.11 Ice Bed

BASES BACKGROUND The ice bed consists of over 2,158,000 lbs of ice stored in 1944 baskets within the ice condenser. Its primary purpose is to provide a large heat

sink in the event of a release of energy from a Design Basis Accident (DBA) in containment. The ice would absorb energy and limit

containment peak pressure and temperature during the accident

transient. Limiting the pressure and temperature reduces the release of

fission product radioactivity from containment to the environment in the

event of a DBA.

The ice condenser is an annular compartment enclosing approximately

300 of the perimeter of the upper containment compartment, but penetrating the operating deck so that a portion extends into the lower

containment compartment. The lower portion has a series of hinged

doors exposed to the atmosphere of the lower containment compartment, which, for normal plant operation, are designed to remain closed. At the

top of the ice condenser is another set of doors exposed to the

atmosphere of the upper compartment, which also remain closed during

normal plant operation. Intermediate deck doors, located below the top

deck doors, form the floor of a plenum at the upper part of the ice

condenser. These doors also remain closed during normal plant

operation. The upper plenum area is used to facilitate surveillance and

maintenance of the ice bed.

The ice baskets contain the ice within the ice condenser. The ice bed is

considered to consist of the total volume from the bottom elevation of the

ice baskets to the top elevation of the ice baskets. The ice baskets

position the ice within the ice bed in an arrangement to promote heat

transfer from steam to ice. This arrangement enhances the ice

condenser's primary function of condensing steam and absorbing heat

energy released to the containment during a DBA.

Ice Bed B 3.6.11 BASES (continued)

Watts Bar - Unit 2 B 3.6-60 (developmental)

A BACKGROUND (continued)

In the event of a DBA, the ice condenser inlet doors (located below the operating deck) open due to the pressure rise in the lower compartment.

This allows air and steam to flow from the lower compartment into the ice

condenser. The resulting pressure increase within the ice condenser

causes the intermediate deck doors and the top deck doors to open, which allows the air to flow out of the ice condenser into the upper

compartment. Steam condensation within the ice condenser limits the

pressure and temperature buildup in containment. A divider barrier

separates the upper and lower compartments and ensures that the steam

is directed into the ice condenser.

The ice, together with the containment spray, is adequate to absorb the

initial blowdown of steam and water from a DBA and the additional heat

loads that would enter containment during several hours following the

initial blowdown. The additional heat loads would come from the residual

heat in the reactor core, the hot piping and components, and the

secondary system, including the steam generators. During the post

blowdown period, the Air Return System (ARS) returns upper

compartment air through the divider barrier to the lower compartment.

This serves to equalize pressures in containment and to continue

circulating heated air and steam from the lower compartment through the

ice condenser where the heat is removed by the remaining ice.

As ice melts, the water passes through the ice condenser floor drains into

the lower compartment. Thus, a second function of the ice bed is to be a

large source of borated water (via the containment sump) for long term

Emergency Core Cooling System (ECCS) and Containment Spray

System heat removal functions in the recirculation mode.

A third function of the ice bed and melted ice is to remove fission product

iodine that may be released from the core during a DBA. Iodine removal

occurs during the ice melt phase of the accident and continues as the

melted ice is sprayed into the containment atmosphere by the

Containment Spray System. The ice is adjusted to an alkaline pH that

facilitates removal of radioactive iodine from the containment atmosphere.

The alkaline pH also minimizes the occurrence of the chloride and

caustic stress corrosion on mechanical systems and components

exposed to ECCS and Containment Spray System fluids in the

recirculation mode of operation.

It is important for the ice to be uniformly distributed around the 24 ice

condenser bays and for open flow paths to exist around ice baskets. This

is especially important during the initial blowdown so that the steam and Ice Bed B 3.6.11 BASES (continued)

Watts Bar - Unit 2 B 3.6-61 (developmental)

A BACKGROUND (continued) water mixture entering the lower compartment do not pass through only part of the ice condenser, depleting the ice there while bypassing the ice

in other bays.

Two phenomena that can degrade the ice bed during the long service

period are:

a. Loss of ice by melting or sublimation; and

b. Obstruction of flow passages through the ice bed due to buildup of frost or ice. Both of these degrading phenomena are reduced by

minimizing air leakage into and out of the ice condenser.

The ice bed limits the temperature and pressure that could be expected

following a DBA, thus limiting leakage of fission product radioactivity from

containment to the environment.

APPLICABLE

SAFETY ANALYSES The limiting DBAs considered relative to containment temperature and

pressure are the loss of coolant accident (LOCA) and the steam line

break (SLB). The LOCA and SLB are analyzed using computer codes

designed to predict the resultant containment pressure and temperature

transients. DBAs are not assumed to occur simultaneously or

consecutively.

Although the ice condenser is a passive system that requires no electrical

power to perform its function, the Containment Spray System and the

ARS also function to assist the ice bed in limiting pressures and

temperatures. Therefore, the postulated DBAs are analyzed in regards to

containment Engineered Safety Feature (ESF) systems, assuming the

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

results in one train each of the Containment Spray System and ARS

being inoperable.

The limiting DBA analyses (Ref. 1) show that the maximum peak

containment pressure results from the LOCA analysis and is calculated to

be less than the containment design pressure. For certain aspects of the

transient accident analyses, maximizing the calculated containment

pressure is not conservative. In particular, the cooling effectiveness of

the ECCS during the core reflood phase of a LOCA analysis increases

with increasing containment backpressure.

Ice Bed B 3.6.11 BASES (continued)

Watts Bar - Unit 2 B 3.6-62 (developmental)

A APPLICABLE SAFETY ANALYSES (continued)

For these calculations, the containment backpressure is calculated in a

manner designed to conservatively minimize, rather than maximize, the

calculated transient containment pressures, in accordance with

10 CFR 50, Appendix K (Ref. 2). The maximum peak containment

atmosphere temperature results from the SLB analysis and is discussed

in the Bases for LCO 3.6.5, "Containment Air Temperature."

In addition to calculating the overall peak containment pressures, the

DBA analyses include calculation of the transient differential pressures

that occur across subcompartment walls during the initial blowdown

phase of the accident transient. The internal containment walls and

structures are designed to withstand these local transient pressure

differentials for the limiting DBAs.

The ice bed satisfies Criterion 3 of the NRC Policy Statement.

LCO The ice bed LCO requires the existence of the required quantity of stored ice, appropriate distribution of the ice and the ice bed, open flow paths

through the ice bed, and appropriate chemical content and pH of the

stored ice. The stored ice functions to absorb heat during a DBA, thereby

limiting containment air temperature and pressure. The chemical content

and pH of the ice provide core SDM (boron content) and remove

radioactive iodine from the containment atmosphere when the melted ice

is recirculated through the ECCS and the Containment Spray System, respectively.

APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the operation of the ice bed.

Therefore, 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, the ice bed is not required to be OPERABLE in

these MODES.

Ice Bed B 3.6.11 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.6-63 (developmental)

B ACTIONS A.1 If the ice bed is inoperable, it must be restored to OPERABLE status

within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The Completion Time was developed based on

operating experience, which confirms that due to the very large mass of

stored ice, the parameters comprising OPERABILITY do not change

appreciably in this time period. Because of this fact, the Surveillance

Frequencies are long (months), except for the ice bed temperature, which

is checked every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . If a degraded condition is identified, even for

temperature, with such a large mass of ice it is not possible for the

degraded condition to significantly degrade further in a 48-hour period.

Therefore, 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is a reasonable amount of time to correct a degraded

condition before initiating a shutdown.

B.1 and B.2

If the ice bed 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . 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

REQUIREMENTS SR 3.6.11.1

Verifying that the maximum temperature of the ice bed is 27 F (value does not account for instrument error) ensures that the ice is kept well below the melting point. The 12-hour Frequency was based on operating

experience, which confirmed that, due to the large mass of stored ice, it is

not possible for the ice bed temperature to degrade significantly within a

12-hour period and was also based on assessing the proximity of the

LCO limit to the melting temperature.

Furthermore, the 12-hour Frequency is considered adequate in view of

indications in the control room, including the alarm, to alert the operator to

an abnormal ice bed temperature condition. This SR may be satisfied by

use of the Ice Bed Temperature Monitoring System.

Ice Bed B 3.6.11 BASES (continued)

Watts Bar - Unit 2 B 3.6-64 (developmental)

A SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.11.2

The weighing program is designed to obtain a representative sample of

the ice baskets. The representative sample shall include 6 baskets from

each of the 24 ice condenser bays and shall consist of one basket from

radial rows 1, 2, 4, 6, 8, and 9. If no basket from a designated row can be

obtained for weighing, a basket from the same row of an adjacent bay

shall be weighed.

The rows chosen include the rows nearest the inside and outside walls of

the ice condenser (rows 1 and 2, and 8 and 9, respectively), where heat

transfer into the ice condenser is most likely to influence melting or

sublimation. Verifying the total weight of ice ensures that there is

adequate ice to absorb the required amount of energy to mitigate the

DBAs.

If a basket is found to contain less than 1100 lb of ice, a representative

sample of 20 additional baskets from the same bay shall be weighed.

The average weight of ice in these 21 baskets (the discrepant basket and

the 20 additional baskets) shall be 1100 lb at a 95% confidence level.

[Value does not account for instrument error.]

Weighing 20 additional baskets from the same bay in the event a

Surveillance reveals that a single basket contains less than 1100 lb

ensures that no local zone exists that is grossly deficient in ice. Such a

zone could experience early melt out during a DBA transient, creating a

path for steam to pass through the ice bed without being condensed. The

Frequency of 18 months was based on ice storage tests and the

allowance built into the required ice mass over and above the mass

assumed in the safety analyses. Operating experience has verified that, with the 18 month Frequency, the weight requirements are maintained

with no significant degradation between surveillances.

SR 3.6.11.3

This SR ensures that the azimuthal distribution of ice is reasonably

uniform, by verifying that the average ice weight in each of three

azimuthal groups of ice condenser bays is within the limit. The

Frequency of 18 months was based on ice storage tests and the

allowance built into the required ice mass over and above the mass

assumed in the safety analyses. Operating experience has verified that, with the 18-month Frequency, the weight requirements are maintained

with no significant degradation between surveillances.

Ice Bed B 3.6.11 BASES (continued)

Watts Bar - Unit 2 B 3.6-65 (developmental)

A SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.11.4

This SR ensures that the air/steam flow channels through the ice bed

have not accumulated ice blockage that exceeds 15 percent of the total

flow area through the ice bed region. The allowable 15 percent buildup of

ice is based on the analysis of the subcompartment response to a design

basis LOCA with partial blockage of the ice bed flow channels. The

analysis did not perform detailed flow area modeling, but rather lumped

the ice condenser bays into six sections ranging from 2.75 bays to

6.5 bays. Individual bays are acceptable with greater than 15 percent

blockage, as long as 15 percent blockage is not exceeded for any

analysis section.

To provide a 95 percent confidence that flow blockage does not exceed

the allowed 15 percent, the visual inspection must be made for at least

54 (33 percent) of the 162 flow channels per ice condenser bay. The

visual inspection of the ice bed flow channels is to inspect the flow area, by looking down from the top of the ice bed, and where view is achievable

up from the bottom of the ice bed. Flow channels to be inspected are

determined by random sample. As the most restrictive flow passage

location is found at a lattice frame elevation, the 15 percent blockage

criteria only applies to "flow channels" that comprise the area:

a. between ice baskets, and

b. past lattice frames and wall panels.

Due to a significantly larger flow area in the regions of the upper deck

grating and the lower inlet plenum and turning vanes, it would require a

gross buildup of ice on these structures to obtain a degradation in

air/steam flow. Therefore, these structures are excluded as part of a flow

channel for application of the 15 percent blockage criteria. Plant and

industry experience have shown that removal of ice from the excluded

structures during the refueling outage is sufficient to ensure they remain

operable throughout the operating cycle. Thus, removal of any gross ice

buildup on the excluded structures is performed following outage

maintenance activities.

Ice Bed B 3.6.11 BASES (continued)

Watts Bar - Unit 2 B 3.6-66 (developmental)

B SURVEILLANCE REQUIREMENTS

SR 3.6.11.4 (continued)

Operating experience has demonstrated that the ice bed is the region that

is the most flow restrictive, due to the normal presence of ice

accumulation on lattice frames and wall panels. The flow area through

the ice basket support platform is not a more restrictive flow area because

it is easily accessible from the lower plenum and is maintained clear of ice

accumulation. There is not a mechanistically credible method for ice to

accumulate on the ice basket support platform during plant operation.

Plant and industry experience has shown that the vertical flow area

through the ice basket support platform remains clear of ice accumulation

that could produce blockage. Normally, only a glaze may develop or exist

on the ice basket support platform which is not significant to blockage of

flow area. Additionally, outage maintenance practices provide measures

to clear the ice basket support platform following maintenance activities of

any accumulation of ice that could block flow areas.

Frost buildup or loose ice is not to be considered as flow channel

blockage, whereas attached ice is considered blockage of a flow channel.

Frost is the solid form of water that is loosely adherent, and can be

brushed off with the open hand.

The Frequency of 18 months was based on ice storage tests and the

allowance built into the required ice mass over and above the mass

assumed in the safety analyses.

SR 3.6.11.5

Verifying the chemical composition of the stored ice ensures that the

stored ice has a boron concentration of 1800 ppm and 2000 ppm as sodium tetraborate and a high pH, 9.0 and 9.5, in order to meet the requirement for borated water when the melted ice is used in the ECCS

recirculation mode of operation. Additionally, the minimum boron

concentration setpoint is used to assure reactor subcriticality in a post

LOCA environment, while the maximum boron concentration is used as

the bounding value in the hot leg switchover timing calculation (Ref. 3).

This is accomplished by obtaining at least 24 ice samples. Each sample

is taken approximately one foot from the top of the ice of each randomly

selected ice basket in each ice condenser bay. The SR is modified by a

NOTE that allows the boron concentration and pH value obtained from

averaging the individual samples' analysis results to satisfy the

requirements of the SR. If either the average boron concentration or the

average pH value is outside their prescribed limit, then entry into ACTION

Condition A is required. Sodium tetraborate has been proven effective in Ice Bed B 3.6.11 BASES (continued)

Watts Bar - Unit 2 B 3.6-67 (developmental)

A SURVEILLANCE REQUIREMENTS

SR 3.6.11.5 (continued)

maintaining the boron content for long storage periods, and it also

enhances the ability of the solution to remove and retain fission product

iodine. The high pH is required to enhance the effectiveness of the ice

and the melted ice in removing iodine from the containment atmosphere.

This pH range also minimizes the occurrence of chloride and caustic

stress corrosion on mechanical systems and components exposed to

ECCS and Containment Spray System fluids in the recirculation mode of

operation. The Frequency of 54 months is intended to be consistent with

the expected length of three fuel cycles, and was developed considering

these facts:

a. Long term ice storage tests have determined that the chemical composition of the stored ice is extremely stable;

b. There are no normal operating mechanisms that decrease the boron concentration of the stored ice, and pH remains within a 9.0 through

9.5 range

when boron concentrations are above approximately

1200 ppm.

c. Operating experience has demonstrated that meeting the boron concentration and pH requirements has never been a problem; and

d. Someone would have to enter the containment to take the sample, and, if the unit is at power, that person would receive a radiation

dose.

SR 3.6.11.6

This SR ensures that a representative sampling of ice baskets, which are

relatively thin walled, perforated cylinders, have not been degraded by

wear, cracks, corrosion, or other damage. Each ice basket must be

raised at least 10 feet for this inspection. However, for baskets where

vertical lifting height is restricted due to overhead obstruction, a camera

shall be used to perform the inspection. The Frequency of 40 months for

a visual inspection of the structural soundness of the ice baskets is based

on engineering judgment and considers such factors as the thickness of

the basket walls relative to corrosion rates expected in their service

environment and the results of the long term ice storage testing.

Ice Bed B 3.6.11 BASES Watts Bar - Unit 2 B 3.6-68 (developmental)

B SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.11.7

This SR ensures that initial ice fill and any subsequent ice additions meet

the boron concentration and pH requirements of SR 3.6.11.5. The SR is

modified by a NOTE that allows the chemical analysis to be performed on

either the liquid or resulting ice of each sodium tetraborate solution

prepared. If ice is obtained from offsite sources, then chemical analysis

data must be obtained for the ice supplied.

REFERENCES 1. Watts Bar FSAR, Section 6.2, "Containment Systems" 2. Title 10, Code of Federal Regulations, Part 50, Appendix K, "ECCS Evaluation Models" 3. Westinghouse Letter, WAT-D-10686, "Upper Limit Ice Boron Concentration In Safety Analysis"

Ice Condenser Doors B 3.6.12 (continued)

Watts Bar - Unit 2 B 3.6-69 (developmental)

A B 3.6 CONTAINMENT SYSTEMS B 3.6.12 Ice Condenser Doors

BASES BACKGROUND The ice condenser doors consist of the inlet doors, the intermediate deck doors, and the top deck doors. The functions of the doors are to:

a. Seal the ice condenser from air leakage during the lifetime of the plant; and

b. Open in the event of a Design Basis Accident (DBA) to direct the hot steam air mixture from the DBA into the ice bed, where the ice would

absorb energy and limit containment peak pressure and temperature

during the accident transient.

Limiting the pressure and temperature following a DBA reduces the

release of fission product radioactivity from containment to the

environment.

The ice condenser is an annular compartment enclosing approximately

300 of the perimeter of the upper containment compartment, but penetrating the operating deck so that a portion extends into the lower

containment compartment. The inlet doors separate the atmosphere of

the lower compartment from the ice bed inside the ice condenser. The

top deck doors are above the ice bed and exposed to the atmosphere of

the upper compartment. The intermediate deck doors, located below the

top deck doors, form the floor of a plenum at the upper part of the ice

condenser. This plenum area is used to facilitate surveillance and

maintenance of the ice bed.

The ice baskets held in the ice bed within the ice condenser are arranged

to promote heat transfer from steam to ice. This arrangement enhances

the ice condenser's primary function of condensing steam and absorbing

heat energy released to the containment during a DBA.

Ice Condenser Doors B 3.6.12 BASES (continued)

Watts Bar - Unit 2 B 3.6-70 (developmental)

A BACKGROUND (continued)

In the event of a DBA, the ice condenser inlet doors (located below the operating deck) open due to the pressure rise in the lower compartment.

This allows air and steam to flow from the lower compartment into the ice

condenser. The resulting pressure increase within the ice condenser

causes the intermediate deck doors and the top deck doors to open, which allows the air to flow out of the ice condenser into the upper

compartment. Steam condensation within the ice condensers limits the

pressure and temperature buildup in containment. A divider barrier

separates the upper and lower compartments and ensures that the steam

is directed into the ice condenser.

The ice, together with the containment spray, serves as a containment

heat removal system and is adequate to absorb the initial blowdown of

steam and water from a DBA as well as the additional heat loads that

would enter containment during the several hours following the initial blowdown. The additional heat loads would come from the residual heat in the reactor core, the hot piping and components, and the secondary

system, including the steam generators. During the post blowdown

period, the Air Return System (ARS) returns upper compartment air

through the divider barrier to the lower compartment. This serves to

equalize pressures in containment and to continue circulating heated air

and steam from the lower compartment through the ice condenser, where

the heat is removed by the remaining ice.

The water from the melted ice drains into the lower compartment where it

serves as a source of borated water (via the containment sump) for the

Emergency Core Cooling System (ECCS) and the Containment Spray

System heat removal functions in the recirculation mode. The ice (via the

Containment Spray System) and the recirculated ice melt also serve to

clean up the containment atmosphere.

The ice condenser doors ensure that the ice stored in the ice bed is

preserved during normal operation (doors closed) and that the ice

condenser functions as designed if called upon to act as a passive heat

sink following a DBA.

APPLICABLE

SAFETY ANALYSES The limiting DBAs considered relative to containment pressure and

temperature are the loss of coolant accident (LOCA) and the steam line

break (SLB). The LOCA and SLB are analyzed using computer codes

designed to predict the resultant containment pressure and temperature

transients. DBAs are assumed not to occur simultaneously or

consecutively.

Ice Condenser Doors B 3.6.12 BASES (continued)

Watts Bar - Unit 2 B 3.6-71 (developmental)

A APPLICABLE SAFETY ANALYSES (continued)

Although the ice condenser is a passive system that requires no electrical

power to perform its function, the Containment Spray System and ARS

also function to assist the ice bed in limiting pressures and temperatures.

Therefore, the postulated DBAs are analyzed with respect to Engineered

Safety Feature (ESF) systems, assuming the loss of one ESF bus, which

is the worst case single active failure and results in one train each of the

Containment Spray System and the ARS being rendered inoperable.

The limiting DBA analyses (Ref. 1) show that the maximum peak

containment pressure results from the LOCA analysis and is calculated to

be less than the containment design pressure. For certain aspects of

transient accident analyses, maximizing the calculated containment

pressure is not conservative. In particular, the cooling effectiveness of

the ECCS during the core reflood phase of a LOCA analysis increases

with increasing containment backpressure. For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the calculated transient

containment pressures, in accordance with 10 CFR 50, Appendix K (Ref. 2).

The maximum peak containment atmosphere temperature results from

the SLB analysis and is discussed in the Bases for LCO 3.6.5, "Containment Air Temperature."

An additional design requirement was imposed on the ice condenser door

design for a small break accident in which the flow of heated air and

steam is not sufficient to fully open the doors.

For this situation, the doors are designed so that all of the doors would

partially open by approximately the same amount. Thus, the partially opened doors would modulate the flow so that each ice bay would receive

an approximately equal fraction of the total flow.

This design feature ensures that the heated air and steam will not flow

preferentially to some ice bays and deplete the ice there without utilizing

the ice in the other bays.

Ice Condenser Doors B 3.6.12 BASES (continued)

Watts Bar - Unit 2 B 3.6-72 (developmental)

A APPLICABLE SAFETY ANALYSES (continued)

In addition to calculating the overall peak containment pressures, the

DBA analyses include the calculation of the transient differential

pressures that would occur across subcompartment walls during the initial

blowdown phase of the accident transient. The internal containment walls

and structures are designed to withstand the local transient pressure

differentials for the limiting DBAs.

The ice condenser doors satisfy Criterion 3 of the NRC Policy Statement.

LCO This LCO establishes the minimum equipment requirements to assure that the ice condenser doors perform their safety function. The ice

condenser inlet doors, intermediate deck doors, and top deck doors must

be closed to minimize air leakage into and out of the ice condenser, with

its attendant leakage of heat into the ice condenser and loss of ice

through melting and sublimation. The doors must be OPERABLE to

ensure the proper opening of the ice condenser in the event of a DBA.

OPERABILITY includes being free of any obstructions that would limit

their opening, and for the inlet doors, being adjusted such that the

opening and closing torques are within limits. The ice condenser doors function with the ice condenser to limit the pressure and temperature that could be expected following a DBA.

APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the operation of the ice condenser

doors. Therefore, the LCO is applicable in MODES 1, 2, 3, and 4.

The probability and consequences of these events in MODES 5 and 6 are

reduced due to the pressure and temperature limitations of these

MODES. Therefore, the ice condenser doors are not required to be

OPERABLE in these MODES.

Ice Condenser Doors B 3.6.12 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.6-73 (developmental)

A ACTIONS A Note provides clarification that, for this LCO, separate Condition entry is allowed for each ice condenser door.

A.1 If one or more ice condenser inlet doors are inoperable due to being

physically restrained from opening, the door(s) must be restored to

OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The Required Action is necessary to

return operation to within the bounds of the containment analysis. The

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is consistent with the ACTIONS of LCO 3.6.1, "Containment," which requires containment to be restored to OPERABLE

status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

B.1 and B.2

If one or more ice condenser doors are determined to be partially open or

otherwise inoperable for reasons other than Condition A or if a door is

found that is not closed, it is acceptable to continue plant operation for up to 14 days, provided the ice bed temperature instrumentation is monitored

once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to ensure that the open or inoperable door is not

allowing enough air leakage to cause the maximum ice bed temperature

to approach the melting point. The Frequency of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is based on the

fact that temperature changes cannot occur rapidly in the ice bed

because of the large mass of ice involved. The 14-day Completion Time

is based on long term ice storage tests that indicate that if the temperature is maintained below 27 F, there would not be a significant loss of ice from sublimation. If the maximum ice bed temperature is > 27 F at any time, or ice bed temperature is not verified to be within the specified Frequency as augmented by the provisions of SR 3.0.2, the

situation reverts to Condition C and a Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is

allowed to restore the inoperable door to OPERABLE status or enter into

Required Actions D.1 and D.2. [NOTE: Entry into Condition B is not

required due to personnel standing on or opening an intermediate deck or

upper deck door for short durations to perform required surveillances, minor maintenance such as ice removal, or routine tasks such as system

walkdowns.]

Ice Condenser Doors B 3.6.12 BASES (continued)

Watts Bar - Unit 2 B 3.6-74 (developmental)

A ACTIONS (continued)

C.1 If Required Actions or Completion Times of B.1 or B.2 are not met, the

doors must be restored to OPERABLE status and closed positions within

48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The 48-hour Completion Time is based on the fact that, with

the very large mass of ice involved, it would not be possible for the

temperature to decrease to the melting point and a significant amount of

ice to melt in a 48-hour period.

D.1 and D.2

If the ice condenser doors 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . 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

REQUIREMENTS SR 3.6.12.1

Verifying, by means of the Inlet Door Position Monitoring System, that the inlet doors are in their closed positions makes the operator aware of an

inadvertent opening of one or more doors. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

ensures that operators on each shift are aware of the status of the doors.

SR 3.6.12.2

Verifying, by visual inspection, that each intermediate deck door is closed

and not impaired by ice, frost, or debris provides assurance that the

intermediate deck doors (which form the floor of the upper plenum where

frequent maintenance on the ice bed is performed) have not been left

open or obstructed. The Frequency of 7 days is based on engineering

judgment and takes into consideration such factors as the frequency of

entry into the intermediate ice condenser deck, the time required for

significant frost buildup, and the probability that a DBA will occur.

Ice Condenser Doors B 3.6.12 BASES (continued)

Watts Bar - Unit 2 B 3.6-75 (developmental)

A SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.12.3

Verifying, by visual inspection, that the ice condenser inlet doors are not

impaired by ice, frost, or debris provides assurance that the doors are

free to open in the event of a DBA. For this unit, the Frequency of

18 months (3 months during the first year after receipt of license - the

3 month performances during the first year after receipt of license may be

extended to coincide with plant outages) is based on door design, which

does not allow water condensation to freeze, and operating experience, which indicates that the inlet doors very rarely fail to meet their SR

acceptance criteria. Because of high radiation in the vicinity of the inlet doors during power operation, this Surveillance is normally performed during a shutdown.

SR 3.6.12.4

Verifying the opening torque of the inlet doors provides assurance that no

doors have become stuck in the closed position. The value of 675 in-lb is

based on the design opening pressure on the doors of 1.0 lb/ft

2. For this unit, the Frequency of 18 months (3 months during the first year after

receipt of license - the 3 month performances during the first year after

receipt of license may be extended to coincide with plant outages) is

based on the passive nature of the closing mechanism (i.e., once

adjusted, there are no known factors that would change the setting, except possibly a buildup of ice; ice buildup is not likely, however, because of the door design, which does not allow water condensation to

freeze). Operating experience indicates that the inlet doors usually meet

their SR acceptance criteria. Because of high radiation in the vicinity of

the inlet doors during power operation, this Surveillance is normally

performed during a shutdown.

Ice Condenser Doors B 3.6.12 BASES (continued)

Watts Bar - Unit 2 B 3.6-76 (developmental)

A SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.12.5

The torque test Surveillance ensures that the inlet doors have not

developed excessive friction and that the return springs are producing a

door return torque within limits. The torque test consists of the following:

1. Verify that the torque, T(OPEN), required to cause opening motion at the 40 open position is 195 in-lb;

2. Verify that the torque, T(CLOSE), required to hold the door stationary (i.e., keep it from closing) at the 40 open position is 78 in-lb; and

3. Calculate the frictional torque,

T(FRICT) = 0.5 {T(OPEN) - T(CLOSE)},

and verify that the T(FRICT) is 40 in-lb.

The purpose of the friction and return torque Specifications is to ensure

that, in the event of a small break LOCA or SLB, all of the 24 door pairs

open uniformly. This assures that, during the initial blowdown phase, the

steam and water mixture entering the lower compartment does not pass

through part of the ice condenser, depleting the ice there, while bypassing

the ice in other bays. The Frequency of 18 months (3 months during the

first year after receipt of license - the 3 month performances during the

first year after receipt of license may be extended to concide with plant

outages) is based on the passive nature of the closing mechanism (i.e.,

once adjusted, there are no known factors that would change the setting, except possibly a buildup of ice; ice buildup is not likely, however, because of the door design, which does not allow water condensation to

freeze). Operating experience indicates that the inlet doors very rarely

fail to meet their SR acceptance criteria. Because of high radiation in the

vicinity of the inlet doors during power operation, this Surveillance is normally performed during a shutdown.

Ice Condenser Doors B 3.6.12 BASES (continued)

Watts Bar - Unit 2 B 3.6-77 (developmental)

A SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.12.6

Verifying the OPERABILITY of the intermediate deck doors provides

assurance that the intermediate deck doors are free to open in the event

of a DBA. The verification consists of visually inspecting the intermediate

doors for structural deterioration, verifying free movement of the vent

assemblies, and ascertaining free movement of each door when lifted

with the applicable force shown below:

DOOR LIFTING FORCE a. Adjacent to crane wall

< 37.4 lb

b. Paired with door adjacent to crane wall 33.8 lb c. Adjacent to containment wall 31.8 lb d. Paired with door adjacent to containment wall 31.0 lb The above test lifting forces were established based upon test results

gathered on newly manufactured Intermediate Deck Doors set up in

fixturing to simulate plant installation tolerances. The lifting force values

developed were to account for and envelope expected door panel

variations in weight and hinge friction and alignments. The intent of the

surveillance is to establish a method of detecting abnormalities or

deteriorating conditions of the door panels or hinges after completion of

refueling outage maintenance activities.

The 18-month Frequency (3 months during the first year after receipt of

license) is based on the passive design of the intermediate deck doors, the frequency of personnel entry into the intermediate deck, and the fact that SR 3.6.12.2 confirms on a 7 day Frequency that the doors are not impaired by ice, frost, or debris, which are ways a door would fail the

opening force test (i.e., by sticking or from increased door weight).

Ice Condenser Doors B 3.6.12 BASES Watts Bar - Unit 2 B 3.6-78 (developmental)

A SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.12.7

Verifying, by visual inspection, that the top deck doors are in place, not

obstructed, and verifying free movement of the vent assembly provides

assurance that the doors are performing their function of keeping warm

air out of the ice condenser during normal operation, and would not be

obstructed if called upon to open in response to a DBA. The Frequency of

92 days is based on engineering judgment, which considered such

factors as the following:

a. The relative inaccessibility and lack of traffic in the vicinity of the doors make it unlikely that a door would be inadvertently left open;

b. Excessive air leakage would be detected by temperature monitoring in the ice condenser; and

c. The light construction of the doors would ensure that, in the event of a DBA, air and gases passing through the ice condenser would find a

flow path, even if a door were obstructed.

REFERENCES

1. Watts Bar FSAR, Section 15.0, "Accident Analysis." 2. Title 10, Code of Federal Regulations, Part 50, Appendix K, "ECCS Evaluation Models."

Divider Barrier Integrity B 3.6.13 (continued)

Watts Bar - Unit 2 B 3.6-79 (developmental)

A B 3.6 CONTAINMENT SYSTEMS B 3.6.13 Divider Barrier Integrity

BASES BACKGROUND The divider barrier consists of the operating deck and associated seals, personnel access doors, and equipment hatches that separate the upper

and lower containment compartments. Divider barrier integrity is

necessary to minimize bypassing of the ice condenser by the hot steam

and air mixture released into the lower compartment during a Design

Basis Accident (DBA). This ensures that most of the gases pass through

the ice bed, which condenses the steam and limits pressure and

temperature during the accident transient. Limiting the pressure and

temperature reduces the release of fission product radioactivity from

containment to the environment in the event of a DBA.

In the event of a DBA, the ice condenser inlet doors (located below the

operating deck) open due to the pressure rise in the lower compartment.

This allows air and steam to flow from the lower compartment into the ice

condenser. The resulting pressure increase within the ice condenser

causes the intermediate deck doors and the door panels at the top of the

condenser to open, which allows the air to flow out of the ice condenser

into the upper compartment. The ice condenses the steam as it enters, thus limiting the pressure and temperature buildup in containment. The

divider barrier separates the upper and lower compartments and ensures

that the steam is directed into the ice condenser. The ice, together with

the containment spray, is adequate to absorb the initial blowdown of

steam and water from a DBA as well as the additional heat loads that

would enter containment over several hours following the initial

blowdown. The additional heat loads would come from the residual heat

in the reactor core, the hot piping and components, and the secondary

system, including the steam generators. During the post blowdown

period, the Air Return System (ARS) returns upper compartment air

through the divider barrier to the lower compartment. This serves to

equalize pressures in containment and to continue circulating heated air

and steam from the lower compartment through the ice condenser, where

the heat is removed by the remaining ice.

Divider barrier integrity ensures that the high energy fluids released

during a DBA would be directed through the ice condenser and that the

ice condenser would function as designed if called upon to act as a

passive heat sink following a DBA.

Divider Barrier Integrity B 3.6.13 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.6-80 (developmental)

A APPLICABLE SAFETY ANALYSES Divider barrier integrity ensures the functioning of the ice condenser to

the limiting containment pressure and temperature that could be

experienced following a DBA. The limiting DBAs considered relative to

containment temperature and pressure are the loss of coolant accident (LOCA) and the steam line break (SLB). The LOCA and SLB are

analyzed using computer codes designed to predict the resultant

containment pressure and temperature transients. DBAs are assumed

not to occur simultaneously or consecutively.

Although the ice condenser is a passive system that requires no electrical

power to perform its function, the Containment Spray System and the

ARS also function to assist the ice bed in limiting pressures and

temperatures. Therefore, the postulated DBAs are analyzed, with respect

to containment Engineered Safety Feature (ESF) systems, assuming the

loss of one ESF bus, which is the worst case single active failure and results in the inoperability of one train in both the Containment Spray System and the ARS.

The limiting DBA analyses (Ref. 1) show that the maximum peak

containment pressure results from the LOCA analysis and is calculated to

be less than the containment design pressure. The maximum peak

containment temperature results from the SLB analysis and is discussed

in the Bases for LCO 3.6.5, "Containment Air Temperature."

In addition to calculating the overall peak containment pressures, the

DBA analyses include calculation of the transient differential pressures

that occur across subcompartment walls during the initial blowdown

phase of the accident transient. The internal containment walls and

structures are designed to withstand these local transient pressure

differentials for the limiting DBAs.

The divider barrier satisfies Criterion 3 of the NRC Policy Statement.

Divider Barrier Integrity B 3.6.13 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.6-81 (developmental)

A LCO This LCO establishes the minimum equipment requirements to ensure that the divider barrier performs its safety function of ensuring that bypass leakage, in the event of a DBA, does not exceed the bypass leakage

assumed in the accident analysis. Included are the requirements that the

personnel access doors and equipment hatches in the divider barrier are

OPERABLE and closed and that the divider barrier seal is properly

installed and has not degraded with time. An exception to the

requirement that the doors be closed is made to allow personnel transit

entry through the divider barrier. The basis of this exception is the

assumption that, for personnel transit, the time during which a door is

open will be short (i.e., shorter than the Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for

Condition A). The divider barrier functions with the ice condenser to limit

the pressure and temperature that could be expected following a DBA.

APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the integrity of the divider barrier.

Therefore, the LCO is applicable in MODES 1, 2, 3, and 4.

The probability and consequences of these events in MODES 5 and 6 are

low due to the pressure and temperature limitations of these MODES. As

such, divider barrier integrity is not required in these MODES.

ACTIONS A.1 If one or more personnel access doors or equipment hatches are

inoperable or open, except for personnel transit entry, 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is allowed to

restore the door(s) and equipment hatches to OPERABLE status and the

closed position. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is consistent with

LCO 3.6.1, "Containment," which requires that containment be restored to

OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Condition A has been modified by a Note to provide clarification that, for

this LCO, separate Condition entry is allowed for each personnel access door or equipment hatch.

Divider Barrier Integrity B 3.6.13 BASES (continued)

Watts Bar - Unit 2 B 3.6-82 (developmental)

A ACTIONS (continued)

B.1 If the divider barrier seal is inoperable, 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is allowed to restore the

seal to OPERABLE status. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is consistent

with LCO 3.6.1, which requires that containment be restored to

OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

C.1 and C.2

If the divider barrier integrity 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . 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

REQUIREMENTS SR 3.6.13.1

Verification, by visual inspection, that all personnel access doors and

equipment hatches between the upper and lower containment

compartments are closed provides assurance that divider barrier integrity

is maintained prior to the reactor being taken from MODE 5 to MODE 4.

The visual inspection shall include the canal gate and control rod drive

missile shield which penetrate the divider barrier. This SR is necessary

because many of the doors and hatches may have been opened for

maintenance during the shutdown.

Divider Barrier Integrity B 3.6.13 BASES (continued)

Watts Bar - Unit 2 B 3.6-83 (developmental)

A SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.13.2

Verification, by visual inspection, that the personnel access door and

equipment hatch seals, sealing surfaces, and alignments are acceptable

provides assurance that divider barrier integrity is maintained. This

inspection cannot be made when the door or hatch is closed. Therefore, SR 3.6.13.2 is required for each door or hatch that has been opened, prior to the final closure. Some doors and hatches may not be opened for

long periods of time. Those that use resilient materials in the seals must

be opened and inspected at least once every 10 years to provide

assurance that the seal material has not aged to the point of degraded performance. The Frequency of 10 years is based on the known resiliency of the materials used for seals, the fact that the openings have

not been opened (to cause wear), and operating experience that confirms

that the seals inspected at this Frequency have been found to be

acceptable.

SR 3.6.13.3

Verification, by visual inspection, after each opening of a personnel

access door or equipment hatch that it has been closed makes the

operator aware of the importance of closing it and thereby provides

additional assurance that divider barrier integrity is maintained while in

applicable MODES.

SR 3.6.13.4

The divider barrier seal can be field spliced for repair purposes utilizing a

cold bond procedure rather than the original field splice technique of

vulcanization. However, the cold bond adhesive, which works in

conjunction with a bolt array to splice the field joint, could not be heat

aged to 40 years plant life prior to acceptability testing. Prolonged

exposure to the elevated temperatures required for heat aging the seal

material was destructive to the adhesive. The seal material was heat

aged to 40 years equivalent age, and the entire joint assembly was

irradiated to 40 year normal operation plus accident integrated dose.

Conducting periodic peel tests on the test specimens provides assurance that the adhesive has not degraded in the containment environment. The Frequencies of 18 months for the first two outages after fabrication of the

joint, followed by 18 months if the peel length is greater than 1/2" and

36 months if the peel length is less than or equal to 1/2" is based upon

the original vendor's recommendation which is based upon baseline

examination of the strength of the adhesive. Therefore, the Frequency

was concluded to be acceptable from a reliability standpoint.

Divider Barrier Integrity B 3.6.13 BASES Watts Bar - Unit 2 B 3.6-84 (developmental)

A SURVEILLANCE REQUIREMENTS

SR 3.6.13.5

Visual inspection of the seal around the perimeter provides assurance

that the seal is properly secured in place. The Frequency of 18 months

was developed considering such factors as the inaccessibility of the seals

and absence of traffic in their vicinity, the strength of the bolts and

mechanisms used to secure the seal, and the plant conditions needed to

perform the SR. Operating experience has shown that these components

usually pass the Surveillance when performed at the 18 month

Frequency. Therefore, the Frequency was concluded to be acceptable

from a reliability standpoint.

REFERENCES

1. Watts Bar FSAR, Section 6.2, "Containment Systems."

Containment Recirculation Drains B 3.6.14 (continued)

Watts Bar - Unit 2 B 3.6-85 (developmental)

A B 3.6 CONTAINMENT SYSTEMS B 3.6.14 Containment Recirculation Drains

BASES BACKGROUND The containment recirculation drai ns consist of the ice condenser drains and the refueling canal drains. The ice condenser is partitioned into

24 bays, each having a pair of inlet doors that open from the bottom

plenum to allow the hot steam-air mixture from a Design Basis Accident (DBA) to enter the ice condenser. Twenty of the 24 bays have an ice

condenser floor drain at the bottom to drain the melted ice into the lower

compartment (in the 4 bays that do not have drains, the water drains

through the floor drains in the adjacent bays). Each drain leads to a drain

pipe that drops down several feet, then makes one or more 90 bends and exits into the lower compartment. A check (flapper) gate at the end of each pipe keeps warm air from entering during normal operation, but when the water exerts pressure, it opens to allow the water to spill into

the lower compartment. This prevents water from backing up and

interfering with the ice condenser inlet doors. The water delivered to the

lower containment serves to cool the atmosphere as it falls through to the

floor and provides a source of borated water at the containment sump for

long term use by the Emergency Core Cooling System (ECCS) and the

Containment Spray System during the recirculation mode of operation.

The two refueling canal drains are at low points in the refueling canal.

During a refueling, plugs are installed in the drains and the canal is

flooded to facilitate the refueling process. The water acts to shield and

cool the spent fuel as it is transferred from the reactor vessel to storage.

After refueling, the canal is drained and the plugs removed. In the event

of a DBA, the refueling canal drains are the main return path to the lower

compartment for Containment Spray Sy stem water sprayed into the upper compartment.

The ice condenser drains and the refueling canal drains function with the

ice bed, the Containment Spray System, and the ECCS to limit the

pressure and temperature that could be expected following a DBA.

Containment Recirculation Drains B 3.6.14 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.6-86 (developmental)

A APPLICABLE SAFETY ANALYSES The limiting DBAs considered relative to containment pressure and

temperature are the loss of coolant accident (LOCA) and the steam line

break (SLB) respectively. The LOCA and SLB are analyzed using

computer codes designed to predict the resultant containment pressure

and temperature transients. DBAs are assumed not to occur

simultaneously or consecutively. Although the ice condenser is a passive

system that requires no electrical power to perform its function, the

Containment Spray System and the Air Return System (ARS) also

function to assist the ice bed in limiting pressures and temperatures.

Therefore, the analysis of the postulated DBAs, with respect to

Engineered Safety Feature (ESF) systems, assumes the loss of one ESF

bus, which is the worst case single active failure and results in one train

of the Containment Spray System and one train of the ARS being

rendered inoperable.

The limiting DBA analyses (Ref. 1) show that the maximum peak containment pressure results from the LOCA analysis and is calculated to

be less than the containment design pressure. The maximum peak

containment atmosphere temperature results from the SLB analysis and

is discussed in the Bases for LCO 3.6.5, "Containment Air Temperature."

In addition to calculating the overall peak containment pressures, the

DBA analyses include calculation of the transient differential pressures

that occur across subcompartment walls during the initial blowdown

phase of the accident transient. The internal containment walls and

structures are designed to withstand these local transient pressure

differentials for the limiting DBAs.

The containment recirculation drains satisfy Criterion 3 of the NRC Policy

Statement.

LCO This LCO establishes the minimum requirements to ensure that the containment recirculation drains perform their safety functions. The ice

condenser floor drain valve gates must be closed to minimize air leakage

into and out of the ice condenser during normal operation and must open

in the event of a DBA when water begins to drain out. The refueling canal

drains must have their plugs removed and remain clear to ensure the

return of Containment Spray System water to the lower containment in

the event of a DBA. The containment recirculation drains function with

the ice condenser, ECCS, and Containment Spray System to limit the

pressure and temperature that could be expected following a DBA.

Containment Recirculation Drains B 3.6.14 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.6-87 (developmental)

A APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature, which would require the operation of the containment recirculation drains. Therefore, the LCO is applicable in

MODES 1, 2, 3, and 4.

The probability and consequences of these events in MODES 5 and 6 are

low due to the pressure and temperature limitations of these MODES. As

such, the containment recirculation drains are not required to be

OPERABLE in these MODES.

ACTIONS A.1 If one ice condenser floor drain is inoperable, 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is allowed to restore

the drain to OPERABLE status. The Required Action is necessary to

return operation to within the bounds of the containment analysis. The

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is consistent with the ACTIONS of LCO 3.6.1, "Containment," which requires that containment be restored to

OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

B.1 If one refueling canal drain is inoperable, 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is allowed to restore the

drain to OPERABLE status. The Required Action is necessary to return

operation to within the bounds of the containment analysis. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

Completion Time is consistent with the ACTIONS of LCO 3.6.1, which

requires that containment be restored to OPERABLE status in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

C.1 and C.2

If the affected drain(s) 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . 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.

Containment Recirculation Drains B 3.6.14 BASES (continued)

Watts Bar - Unit 2 B 3.6-88 (developmental)

A SURVEILLANCE REQUIREMENTS SR 3.6.14.1

Verifying the OPERABILITY of the refueling canal drains ensures that

they will be able to perform their functions in the event of a DBA. This

Surveillance confirms that the refueling canal drain plugs have been

removed and that the drains are clear of any obstructions that could

impair their functioning. In addition to debris near the drains, attention

must be given to any debris that is located where it could be moved to the

drains in the event that the Containment Spray System is in operation and

water is flowing to the drains. SR 3.6.14.1 must be performed before

entering MODE 4 from MODE 5 after every filling of the canal to ensure that the plugs have been removed and that no debris that could impair the drains was deposited during the time the canal was filled. The 92 day

Frequency was developed considering such factors as the inaccessibility

of the drains, the absence of traffic in the vicinity of the drains, and the

redundancy of the drains.

SR 3.6.14.2

Verifying the OPERABILITY of the ice condenser floor drains ensures that

they will be able to perform their functions in the event of a DBA.

Inspecting the drain valve gate ensures that the gate is performing its

function of sealing the drain line from warm air leakage into the ice

condenser during normal operation, yet will open if melted ice fills the line

following a DBA. Verifying that the drain lines are not obstructed ensures

their readiness to drain water from the ice condenser. The 18 month

Frequency was developed considering such factors as the inaccessibility

of the drains during power operation; the design of the ice condenser, which precludes melting and refreezing of the ice; and operating experience that has confirmed that the drains are found to be acceptable

when the Surveillance is performed at an 18 month Frequency. Because

of high radiation in the vicinity of the drains during power operation, this

Surveillance is normally done during a shutdown.

REFERENCES

1. Watts Bar FSAR, Section 6.2, "Containment Systems."

Shield Building B 3.6.15 (continued)

Watts Bar - Unit 2 B 3.6-89 (developmental)

A B 3.6 CONTAINMENT SYSTEMS B 3.6.15 Shield Building

BASES BACKGROUND The shield building is a concrete structure that surrounds the steel containment vessel. Between the containment vessel and the shield

building inner wall is an annular space that collects containment leakage

that may occur following a loss of coolant accident (LOCA) as well as

other design basis accidents (DBAs) that release radioactive material.

This space also allows for periodic inspection of the outer surface of the

steel containment vessel.

The Emergency Gas Treatment System (EGTS) establishes a negative

pressure in the annulus between the shield building and the steel

containment vessel. Filters in the system then control the release of

radioactive contaminants to the environment. The shield building is

required to be OPERABLE to ensure retention of containment leakage

and proper operation of the EGTS.

APPLICABLE

SAFETY ANALYSES The design basis for shield building OPERABILITY is a LOCA.

Maintaining shield building OPERABILITY ensures that the release of

radioactive material from the containment atmosphere is restricted to

those leakage paths and associated leakage rates assumed in the

accident analyses.

The shield building satisfies Criterion 3 of the NRC Policy Statement.

LCO Shield building OPERABILITY must be maintained to ensure proper operation of the EGTS and to limit radioactive leakage from the

containment to those paths and leakage rates assumed in the accident

analyses.

Shield Building B 3.6.15 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.6-90 (developmental)

A APPLICABILITY Maintaining shield building OPERABILITY prevents leakage of radioactive material from the shield building. Radioactive material may enter the shield building from the containment following a DBA. Therefore, shield

building OPERABILITY is required in MODES 1, 2, 3, and 4 when DBAs

could release radioactive material to the containment atmosphere.

In MODES 5 and 6, the probability and consequences of these events are

low due to the Reactor Coolant System temperature and pressure

limitations in these MODES. Therefore, shield building OPERABILITY is

not required in MODE 5 or 6.

ACTIONS A.1

In the event shield building OPERABILITY is not maintained, shield

building OPERABILITY must be restored within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a

reasonable Completion Time considering the limited leakage design of

containment and the low probability of a Design Basis Accident occurring

during this time period.

B.1 The Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is based on engineering judgment. The

normal alignment for both EGTS control loops is the A-Auto position.

With both EGTS control loops in A-Auto, both trains will function upon

initiation of a Containment Isolation Phase A (CIA) signal. In the event of

a LOCA, the annulus vacuum control system isolates and both trains of

the EGTS pressure control loops will be placed in service to maintain the

required negative pressure. If annulus vacuum is lost during normal

operations, the A-Auto position is unaffected by the loss of vacuum. This

operational configuration is acceptable because the accident dose

analysis conservatively assumes the annulus is at atmospheric pressure

at event initiation. A Note has been provided which makes the

requirement to maintain the annulus pressure within limits not applicable

during venting operations, required annulus entries, or Auxiliary Building

isolations not exceeding 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in duration.

Shield Building B 3.6.15 BASES (continued)

Watts Bar - Unit 2 B 3.6-91 (developmental)

B ACTIONS (continued)

C.1 and C.2 If the shield building 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . 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

REQUIREMENTS SR 3.6.15.1

Verifying that shield building annulus negative pressure is within limit (equal to or more negative than -5 inches water gauge; value does not

account for instrument error) ensures that operation remains within the limit assumed in the containment analysis. The 12-hour Frequency of this

SR was developed considering operating experience related to shield

building annulus pressure variations and pressure instrument drift during

the applicable MODES.

SR 3.6.15.2

Maintaining shield building OPERABILITY requires maintaining each door

in the access opening closed, except when the access opening is being

used for normal transient entry and exit. The 31-day Frequency of this

SR is based on engineering judgment and is considered adequate in view

of the other indications of door status that are available to the operator.

SR 3.6.15.3

This SR would give advance indication of gross deterioration of the

concrete structural integrity of the shield building. The Frequency of this

SR is the same as that of SR 3.6.1.1. The verification is done during

shutdown.

Shield Building B 3.6.15 BASES Watts Bar - Unit 2 B 3.6-92 (developmental)

B SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.15.4

The EGTS is required to maintain a pressure equal to or more negative

than -0.50 inches water gauge ("wg) in the annulus at an elevation

equivalent to the top of the Auxiliary Building. At elevations higher than

the Auxiliary Building, the EGTS is required to maintain a pressure equal

to or more negative than -0.25 "wg. The low pressure sense line for the

pressure controller is located in the annulus at elevation 783. By verifying

that the annulus pressure is equal to or more negative than -0.61 "wg at

elevation 783, the annulus pressurization requirements stated above are

met. The ability of a EGTS train with final flow 3600 cfm and 4400 cfm to produce the required negative pressure during the test operation provides assurance that the building is adequately sealed. The

negative pressure prevents leakage from the building, since outside air

will be drawn in by the low pressure at a maximum rate 250 cfm. The 18 month Frequency on a STAGGERED TEST BASIS is consistent with

Regulatory Guide 1.52 (Ref. 1) guidance for functional testing.

REFERENCES 1. Regulatory Guide 1.52, Revision 2, "Design, Testing and Maintenance Criteria for Post Accident Engineered-Safety-Feature

Atmospheric Cleanup System Air Filtration and Adsorption Units of

Light-Water Cooled Nuclear Power Plants."

@@ Line 2: / Line 2: @@
 | number = ML100550590
 | issue date = 02/02/2010
-| title = Watts Bar Nuclear Plant, Unit 2, Developmental Revision B - Technical Specifications Bases B 3.6 - Containment Systems
+| title = Developmental Revision B - Technical Specifications Bases B 3.6 - Containment Systems
 | author name =
 | author affiliation = Tennessee Valley Authority

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