ML20209B773

13.

FSAR, Sections 5.2.7 "RCBP Leakage Detection Systems" and 12.2.4 " Airborne Radioactivity Monitoring."

SEQUOYAH - UNIT 1 B 3/4 4-4e

. Amendment No.

i

REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE BACKGROUND Components that contain or transport the coolant to or from the reactor core make up the reacter coolant system (RCS).

Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from the RCS.

During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant leakage, through either normal operational wear or mechanical deterioration.

The purpose of the RCS Operational leakage LCO is to limit system operation in the presence of leakage from these sources to amounts that do not compromise safety. This LCO specifies the types and amounts of leakage.

10 CFR 50, Appendix A, GDC 30 (Ref. 1), requires means for detecting and, to the extent practical, identifying the source of reactor coolant leakage.

Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.

The safety significance oi RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring reactor coolant leakage into the containment area is necessary. Quickly separating the identified LEAKAGE from the unidentified leakage is necessary to provide quantitative information to the cperators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public.

A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100% leaktight.

Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible,.to

)

not interfere with RCS leakage detection.

l This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded.

The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA).

APPLICABLE Except for primary-to-secondary leakage, the safety analyses SAFETY ANALYSES do not address operational leakage.

However, other operational leakage is related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event.

The safety analysis for an event resulting in steam discharge to the atmosphere assumes a 1 gpm primary to secondary leakage as the initial condition.

SEQUOYAH - UNIT 1 B 3/4 4-4f Amendment No. 36, 189, 214, 222, 237, i

)

\\

l REACTOR COOLANT SYSTEM l

BASES Primary to secondary leakage is a factor in the dose releases outside containment resulting from a steam line break (SLB) accident. To a lesser extent, other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR).

The leakage contaminates the secondary fluid.

The FSAR (Ref. 3) analysis for SGTR assumes the contaminated secondary fluid is released via safety valves for up to 30 minutes.

Operator action is taken to isolate the affected steam generator within this time period. The 1 gpm primary to secondary leakage is relatively inconsequential.

The SLB is more limiting for site radiation releases. The safety analysis for the SLB accident assumes 1 gpm primary to secondary leakage in one generator as an initial condition.

The dose consequences resulting from the SLB l

accident are well within the limits defined in 10 CFR 100 or the staff approved licensing basis (i.e., a small fraction of these limits).

Based on the NDE uncertainties, bobbin coil voltage distribution and crack growth rate from the previous inspection, the expected leak rate following a steam line rupture is limited to below 8.21 gpm at atmospheric conditions and 70*F in the-faulted loop, which R241 will limit the calculated offsite doses to within 10 percent i

of the 10 CFR 100 guidelines.

If the projected and cycle distribution of crack indications results in primary-to-secondary leakage greater than 8.21 gpm in the faulted loop during a postulated steam line break event, additional tubes must be removed from service in order to reduce the

-postulated primary-to-secondary steam line break leakage to below 8.21 gpm.

The RCS operational leakage satisfies Criterion 2 of the NRC Policy Statement.

LCO RCS operational leakage shall be limited to:

a.

PRESSURE BOUNDARY LEAKhqE No PRESSURE BOUNDARY LEAKAGE is allowed, being indicative of material deterioration.

Leakage of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher leakage.

Violation of this LCO could result in continued degradation of the RCPB.

Leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

b.

UNIDENTIFIED LEAKAGE One gpm of UNIDENTIFIED LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment pocket SEQUOYAH - UNIT 1 B 3/4 4-4g Amendment No. 36, 189, 214, 222, 237,

REACTOR COOLANT SYSTEM BASES sump level monitoring equipment can detect within a reasonable. time period. Violation of this LCO could result in continued degradation of the RCPB, if the leakage is from the pressure boundary.

c.

Primary to Secondary Leakage through Any One Steam Generator (SG)

The 150 gallons per day limit on one SG is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line rupture.

If leaked through many cracks, the cracks are very small, and the above assumption is conservative.

The 150-gallons per day limit incorporated into Surveillance 4.4.6.2.1 is more restrictive than the standard operating leakage limit and is intended to-provide an additional margin to accommodate a crack which might grow at a greater than expected rate or R241 unexpectedly extend outside the thickness of the tube support plate. Hence, the reduced leakage limit, when combined with an effective leak rate monitoring prcgram, provides' additional assurance that, should a significant leak be experienced, it will be detected, and the plant shut down in a timely manner.

d.

IDENTIFIED LEAKAGE Up to 10 gpm of IDENTIFIED LEAKAGE is considered allowable because leakage is from known sources that do not interfere with detection of UNIDENTIFIED LEAKAGE and is well within the capability of the RCS Makeup _ System.

IDENTIFIED LEAKAGE includes leakage to the containment from specifically known and located sources, but does not include PRESSURE BOUNDARY LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered leakage).

Violation of this LCO could result in continued degradation of a component or system.

APPLICABILITY In MODES 1, 2, 3, and 4, the potential for reactor coolant PRESSURE BOUNDARY LEAKAGE is greatest when the RCS is pressurized.

In MODES 5 and 6, leakage limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for leakage.

SEQUOYAH -~ UNIT 1 B 3/4 4-4h Amendment No. 36, 189, 214, 222, 237,

E

(

[

REACTOR COOLANT SYST @

BASES LCO 3/4.4.6.3, "RCS Pressure Isolation Valve (PIV) Leakage,"

. measures leakage through each individual PIV and can impact this LCO.

Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS leakage when the other is leak tight.

If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable IDENTIFIED LEAKAGE.

i ACTIONS.

Action a:

i l

If'any PRESSURE BOUNDARY LEAKAGE exists, the reactor must be l

brought to lower pressure conditions to reduce the severity of the leakage and its potential consequences.

It should be noted that leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

The' reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within_the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

This action reduces the leakage and also reduces the factors that-tend to degrade the pressure boundary.

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.

In MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.

Action b:

I UNIDENTIFIED LEAKAGE, IDENTIFIED LEAKAGE, or primary-to-l secondary: leakage in excess.of the LCO limits must be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

This completion time allows' time to verify leakage rates and either identify UNIDENTIFIED LEAKAGE or reduce leakage to within limits before the reactor must be shut down.

This action is

(

necessary to prevent further' deterioration of the RCPB.

If H

UNIDENTIFIED LEAKAGE, IDENTIFIED LEAKAGE, or primary to secondary leakage cannot be reduced to within limits within l.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the reactor must be brought to lower pressure l

conditions to reduce the severity of the leakage and its potential consequences.

The reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

j This action reduces the leakage and also reduces the factors j

that tend to degrade the pressure boundary.

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 1

challenging plant systems.

In MODE 5, the pressure stresses I

acting on the_RCPB are much lower, and further deterioration is much less likely.

-SEQUOYAH - UNIT 1 B 3/4 4-41 Amendment No. 36, 189, 214, 222, 237,

l l

l REACTOR COOLANT SYSTEM BASES i

l I

SURVEILLANCE Surveillance 4.4.6.2.1 REQUIREMENTS Verifying RCS leakage to be within the LCO limits ensures the integrity of the RCPB is maintained.

PRESSURE BOUNDARY LEAKAGE would at first appear as UNIDENTIFIED LEAKAGE and can only be positively identified by inspection.

It should

]

be noted that leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

UNIDENTIFIED LEAKAGE and IDENTIFIED LEAKAGE are determined by performance of an RCS water inventory balance.

Pricary-to-secondary leakage is also measured by performance of an RCS water inventory balance in conjunction with effluent monitoring within the secondary steam and feedwater systems.

The RCS water inventory balance must be met with the reactor at steady state operating conditions (stable pressure, temperature, power level, pressurizer and makeup tank levels, makeup, letdown, and RCP seal injection and return flows).

Therefore, a footnote is added allowing that this i

SR is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady state operation. The 12-hour allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

Performance of this surveillance within the 12-hour allowance is required to maintain compliance with the provisions of Specification 4.0.3.

Steady state operation'is required to perform a proper inventory balance since calculations during maneuvering are not useful.

For RCS operational leakage determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

An early warning of PRESSURE BOUNDARY LEAKAGE or UNIDENTIFIED LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment pocket sump level.

It should be noted that leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE. These leakage detection systems are specified in LCO 3/4.4.6.1, " Leakage Detection Instrumentation."

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months frequency is a reasonable interval to trend leakage and recognizes the importance of early leakage detection in.the prevention of accidents.

Surveillance 4.4.6.2.2 This surveillance provides the means necessary to determine SG OPERABILITY in an operational MODE.

The requirement to i

demonstrate SG tube integrity in accordance with the Steam l

Generator Tube Surveillance Program emphasizes the importance of SG tube integrity, even though this surveillance cannot be performed at normal operating conditions.

SEQUOYAH - UNIT 1 B 3/4 4-4j Amendment No. 36, 189, 214, 222, 237,

REACTOR COOLANT SYSTEM BASES REFERENCES 1.

2.

Regulatory Guide 1.45, May 1973.

3.

FSAR, Section 15.4.3.

I SEQUOYAH - UNIT 1 B 3/4 4-4k Amendment No. 36, 189, 214, 222, 237,

REACTOR COOLANT SYSTEM BASES 3/4.4.6.3 REACTOR COOLANT' SYSTEM PRESSURE ISOLATION VALVE LEAKAGE BACKGROUND 10 CFR 50.2,.10 CFR 50.55a(c), and GDC 55 of 10 CFR 50, Appendix A (Refs. 1, 2, and 3), define reactor coolant system (RCS) pressure isolation valves (PIVs) as any two normally closed valves in series l

within the reactor coolant-pressure boundary (RCPB),

1 which separate the high pressure RCS from an attached low pressure system.

During their lives, these valves can produce varying amounts of reactor coolant leakage through either-normal operational wear or mechanical deterioration.

The RCS_PIV leakage LCO allows RCS high pressure operation when leakage through these valves exists in amounts that do not compromise safety.

Although this specification provides a limit on allowable PIV leakage rate, its main purpose is to prevent overpressure failure of the low pressure portions of connecting systems.

The leakage limit is an indication that the PIVs between the RCS and the connecting systems are degraded or degrading.

PIV leakage could lead to overpressure of the low pressure piping or components.

Failure consequences could be a loss of coolant accident (LOCA) outside of containment or an unanalyzed accident that could degrade the ability for low pressure injection.

The basis for this LCO is the 1975 NRC, " Reactor Safety Study," (Ref. 4) that identified potential intersystem LOCAs as a significant contributor to the risk of core melt.

A subsequent study (Ref. 5) evaluated various PIV configurations to determine the probability of intersystem LOCAs.

PIVs are provided to isolate the RCS from the following typically connected systems:

a.

Residual Heat Removal (RER) System; b.

Safety Injection System; and c.

Chemical and Volume Control System.

The PIVs are listed in Table 3.4-1.

Violation of this LCO could result in continued degradation of a PIV, which could lead to overpressnrization of a low pressure system and the loss of.the integrity of a fission product barrier.

l SEQUOYAH UNIT 1 B 3/4 4-41 Amendment No.

k

REACTOR COOLANT SYSTEM BASES APPLICABLE Reference 4 identified potential intersystem LOCAs as SAFETY ANALYSES a significant contributor to the risk of core melt.

The dominant accident sequence in the intersystem LOCA category is the failure of the low pressure portion of the RHR System outside of containment. The accident is the result of a postulated failure of the PIVs, which are part of the RCPB, and the subsequent pressurization of the RHR System downstream of the PIVs from the RCS.

Because the low pressure portion of the RHR System is typically designed for 600 psig, overpressurization failure of the RHR low pressure line would result in a LOCA outside containment and subsequent risk of core melt.

Reference 5 evaluated various PIV configurations, leakage testing of the valves, and operational changes to determine the effect on the probability of intersystem LOCAs. This study concluded that periodic leakage testing of the PIVs can substantially reduce the probability of an intersystem LOCA.

RCS PIV leakage satisfies Criterion 2 of the NRC Policy Statement.

LCO RCS PIV leakage is IDENTIFIED LEAKAGE into closed systems connected to the RCS.

Isolation valve leakage is usually on the order of drops per minute.

Leakage that increases significantly suggests that something j

is operationally wrong and corrective action must be taken.

The LCO PIV leakage limit is 0.5 gpm per nominal inch of valve size with a maximum limit of 5 gpm.

The previous criterion of 1 gpm for all valve sizes imposed an unjustified penalty on the larger valves without providing information on potential valve degradation and resulted in higher personnel radiation exposures. A study concluded a leakage rate limit based on valve size was superior to a single allowable value.

Reference 6 permits leakage testing at a lower pressure differential than between the specified maximum RCS pressure and the normal' pressure of the connected system during RCS operation (the maximum pressure differential) in those types of valves in which the higher service pressure will tend to diminish the overall leakage channel opening.

In such cases, the observed rate may be adjusted to the maximum pressure differential by assuming leakage is directly proportional to the pressure differential to i

the one half power.

SEQUOYAH - UNIT 1 B 3/4 4-4m Amendment No.

l REACTOR COOLANT SYSTEM

.-BASES APPLICABILITY In MODES 1, 2,

3, and 4, this LCO applies because the PIV leakage potential is greatest when the RCS is pressurized.

In MODE 4, valves in the RHR flow path are not required to meet the requirements of this LCO when in,' or during the transition to or from, the RHR mode of operation.

In MODES 5 and 6, leakage limits are not provided because the lower reactor coolant pressure results in a reduced potential for leakage and for a LOCA outside the containment.

ACTIONS Action a:

The flow path must be isolated. Action a is modified i

by a note that the valves used for isolation must meet the same leakage requirements as the PIVs and must be within the RCPB.

Action a requires that the isolation with one valve must be performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Four hours provides time to reduce leakage in excess of the allowable limit and to isolate the affected system if leakage cannot be reduced.

The 4-hour completion time allows the actions and restricts the operation with leaking isolation valves.

The 72-hour completion time after exceeding the limit allows for the restoration of the leaking PIV to OPERABLE status. This timeframe considers the time required to complete this action and the low probability of a second valve failing during this period.

If leakage cannot be reduced or the system isolated, the plant must be brought to a MODE in which the requirement does not apply.

To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE S within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

This Action may reduce the leakage and also reduces the potential for a LOCA outside the containment.

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.

Action b:

Action b provides clarification that each flow path allows-separate entry into Action a This is allowed basea upon the functional independence of the flow path.

.SEQUOYAH - UNIT 1 B 3/4 4-4n Amendment No.

REACTOR COOLANT SYSTEM BASES Action c:

Action c requires an evaluation of affected systems if a PIV is inoperable. The leakage may have affected system operability or isolation of a leaking flow path with an alternate valve may have degraded the ability of the interconne:ted system to perform its safety function.

SURVEILLANCE Surveillance 4.4.6.3 REQUIREMENTS Performance of leakage testing on'each RCS PIV or isolation valve used to satisfy Action a is required to verify that leakage-is below the specified limi:

and to idantify each leaking valve. The leakage limit of 0.5 gpm per inch of nominal valve diameter up to 5 gpm maximum applies to each valve.

Leakage testing requires a stable pressure condition.

(~

For the two PIVs in series, the leakage requirement l

applies to each valve individually and not to the l

combined leakage across both valves.

If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the other l

valve in series meets the leakage requirement.

In this situation, the protection provided by redundant l

valves would be lost.

l Testing is to be performed every 18 months, a typical refueling cycle, if the plant does not go into MODE 5 for at least 7 days. The 18 month frequency is consistent with 10 CFR 50.55a(g) (Ref. 7) as contained in the Inservice Testing Program, is within frequency allowed by the American Society of Mechanical Engineers (ASME). Code,Section XI (Ref. 6), and is based on the need to perform such surveillances under the conditions that apply during an outage and the potential-for an unplanned transient if the surveillances were performed with the reactor at power.

In addition, testing must be performed once after the valve has been opened by flow or exercised to ensure tight. reseating.

PIVs disturbed in.the performance of this surveillance should also be tested unless documentation shows that an infinite testing loop l

cannot practically be avoided.

Testing must be l

-performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the valve has been l

reseated. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a reasonable and practical time limit for performing this test after opening or reseating a valve.

t l-

-SEQUOYAH - UNIT 1 B 3/4 4-4o Amendment No.

l 1

l

REACTOR COOLANT SYSTEM BASES The leakage limit is to be met at the RCS pressure associated with MODES 1 and 2.

This permits leakage

-testing at high differential pressures with stable conditions not possible in the MODES with lower pressures.

Entry into MODES 3 and 4 is allowed to establish the necessary differential pressures and stable conditions to allow for performance of this surveillance. The note that allows this provision is complementary to the frequency of prior to entry into MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months.

In addition, this surveillance is not required to be performed on the RHR System when the RHR System is aligned to the RCS in the shutdown cooling mode of operation.

PIVs contained in the RHR shutdown cooling flow path must be leakage rate tested after RHR is secured and stable unit conditions and the necessary differential pressures are established.

REFERENCES 1.

2.

10 CFR 50.55a(c).

3.

10 CFR 50, Appendix A, Section V, GDC 55.

4.

WASH-1400 (NUREG-75/014), Appendix V, October 1975.

5.

NUREG-0677, May 1980.

6.

ASME, Boiler and Pressure Vessel Code,Section XI.

7.

EMERGENCY CORE COOLING SYSTEM l

SEQUOYAH - UNIT 1 B 3/4 4-4p Amendment No.

REACTOR C0OLANT SYSTD4

(?Q

')

(...;

3/4.4.7 CHDtISTRY

%:c:

The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady state Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant system over the life of the plant. The associated effects of exceeding the oxygen, chloride and fluoride limits are time and tempersture dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady-State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval pemitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

f

\\

Vs k.

4 i

SEQUOYMi - UNIT 1 B 3/4 4-4 gndmenth.36,109,

),

,3 5

October 11, 1995

I l

I

)

l EMERGENCY CORE COOLING SYSTEM f

BASES 3/4.5.6 SEAL' INJECTION FLOW BACKGROUND The function of the seal injection throttle valves during an accident is similar to the function of the ECCS throttle valves in that each restricts flow from the centrifugal charging pump header to the Reactor Coolant System (RCS).

Theo restriction on reactor coolant pump (RCP) seal injection flow limits the amount of ECCS flow that would be diverted from the injection path following an accident. This limit is based on safety analysis i

assumptions that are required because RCP seal j

injection flow is not isolated during safety injection.

APPLICABLE All ECCS subsystems are taken credit for in the large SAFETY ANALYSES break loss of coolant accident (LOCA) at full power (Ref. 1).

The LOCA analysis establishes the minimum flow for the ECCS pumps. The centrifugal charging pumps are also credited in the small break LOCA analysis. This analysis establishes the flow and discharge head at the design point for the centrifugal charging pumps.

The steam generator tube rupture and main steam line break event analyses also credit the i

centrifugal charging pumps, but are not limiting in

)

their design.

Reference to these analyses is made in assessing changes to the Seal Injection System for i

i evaluation of their effects in relation to the acceptance limits in these analyses.

This LCO ensures that seal injection flow will be sufficient: for RCP seal integrity but limited so that

'the ECCS trains will be capable of delivering i

sufficient _ water to match boiloff rates soon enough to minimize uncovering of the core following a large LOCA.

It.also ensures that the centrifugal charging pumps will deliver sufficient water for a small LOCA i

and sufficient boron to maintain the core suberitical.

l For smaller LOCAs, the charging pumps alone deliver sufficient fluid to overcome the loss and maintain RCS i

inventory.

Seal-injection flow satisfies Criterion 2 of the NRC Policy Statement.

SEQUOYAH - UNIT 1 B 3/4 5-4 Amendment No.

F 1

l DASES f

LCO The intent of the LCO limit on seal injection flow is to make sure that flow through the RCP seal water I

injection line is low enough to ensure that sufficient centrifugal charging pump injection flow is directed to the RCS via the injection points (Ref. 2),

j l

The LCO is not strictly a flow limit, but rather a i

flow limit based on a flow line resistance.

In order 4

to establish the proper flow line resistance, a j

pressure and flow must be known.

The flow line a

resistance is established by adjusting the RCP seal I

injection needle valves to provide a total seal injection flow in the acceptable region of Figure B 3.5.6-1.

The centrifugal charging pump discharge header pressure remains essentially constant j

through all the applicable MODES of this LCO.

A reduction in RCS pressure would result in more flow i

being diverted to the RCP seal injection line than at normal operating pressure.

The valve settings established at the prescribed centrifugal charging j

pump discharge header pressure result in a conservative valve position should RCS pressure decrease.

The flow limits established by Figure B 3.5.6-1 are consistent with the accident l

analysis, j

i The limits on seal injection flow must be met to render the ECCS OPERABLE.

If these conditions are not i

met, the ECCS flow will not be as assumed in the l

accident analyses.

I APPLICABILITY In MODES 1, 2, and 3, the seal injection flow limit is dictated by ECCS flow requirements, which are 1

specified for MODES 1, 2,

3, and 4.

The seal j

injection flow limit is not applicable for MODE 4 and l

lower, however, because high seal injection flow is l

less critical as a result of the lower initial RCS pressure and decay heat removal requirements in these MODES.

Therefore, RCP seal injection flow must be limited in MODES 1, 2,

and 3 to ensure adequate ECCS performance.

]

I l

i f

l t

SEQUOYAH - UNIT 1 B 3/4 5-5 Amendment No.

I

1 1

l' EMERGENCY CORE COOLING SYSTEM BASES ACTION With the seal injection flow exceeding its limit, the amount of charging flow available to the RCS may be reduced.

Under this condition, action must be taken to-restore the flow to below its limit.

The operator has 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from the time the flow is known to be above the limit to correctly position the manual valves and thus be in compliance with the accident analysis.

The' completion time minimizes the potential exposure of the plant to a LOCA with insufficient injection flow and provides a reasonable time to restore seal injection flow within limits. This time is conservative with respect to the completion times of other ECCS LCOs; it is based on operating experience and is sufficient for taking corrective actions by operations personnel.

When the actions cannot be completed within the required completion time, a controlled shutdown must be initiated. The completion time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for reaching MODE 3 from MODE 1 is a reasonable time for a controlled shutdown, based on operating experience and normal cooldown rates,-and does not challenge plant safety systems or operators.

Continuing the plant shutdown from MODE 3, an additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is a reasonable time, based on operating experience and normal cooldown rates, to reach MODE 4, where this LCO is no longer applicable.

SURVEILLANCE Surveillance 4.5.6 REQUIREMENTS Verification every 31 days that the manual seal injection throttle valves are adjusted to give a flow

-within the limit ensures that proper manual seal injection throttle valve position, and hence, proper seal injection flow, is maintained.

The differential pressure that is above the reference minimum value is established between the charging header (PT 62-92) and l

the RCS, and total seal injection flow is verified to be within the limits determined in accordance with the ECCS safety analysis (Ref. 3).

The seal water injection flow limits are shown in Figure B 3.5.6-1.

l The frequency of 31 days is based on engineering judgment and is consistent with other ECCS valve surveillance frequencies. The frequency has proven to be acceptable through operating experience.

The requirements for charging flow vary widely according to plant status and configuration. When charging flow is adjusted, the positions of the air-operated valves, which control charging flow, are

'SEQUOYAH - UNIT 1 B 3/4 5-6 Amendment No.

EMERGENCY CORE COOLING SYSTEM BASES adjusted to balance the flows through the charging header and through the seal injection header to ensure that the seal injection flow to the RCPs is maintained between 8 and 13 gpm per pump.

The reference minimum differential pressure across the seal injection needle valves ensures that regardless of the varied settings of the charging flow control valves that are required to support optimum charging flow, a reference test condition can be established to ensure that flows across the needle valves are within the safety analysis.

The values in the safety analysis for this reference set of conditions are calculated based on conditions during power operation and they are correlated to the minimum ECCS flow to be maintained under the most limiting accident conditions.

As noted, the surveillance is not required to be performed until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the RCS pressure has stabilized within a i 20 psig range of normal operating pressure. The RCS pressure requirement is specified since this configuration will produce the required pressure conditions necessary to assure that the manual valves are set correctly.

The exception is limited to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to ensure that the surveillance is timely.

Performance of this surveillance within the 4-hour allowance is required to maintain compliance with the provisions of Specification 4.0.3.

REFERENCES 1.

FSAR, Chapter 6.3 " Emergency Core Cooling System" and Chapter 15.0 " Accident Analysis".

2.

3.

Westinghouse Electric Company Calculat. ion CN-FSE-99-48 l

SEQUOYAH - UNIT 1 B 3/4 5-7 Amendment No.

[..

lI l

l 1lIl!l1llliiJl11 s,

.oN tne m

d 0

n 5

e m

A

~

5 4

4

.. r

.wD s

t 0

i

)

1 m 4 m

- i p

6. L f

g

(

5. ow EL w

o

6. ABN, l

8 3l F n

vN'. P G..

F c

O.

5 A

B MTI s o

/

n

X' E E.

4 Eo 3

io'. C R it i

c B

Rt yC e

c

?.

Ue

>A j

s In j

i l

Gn

'i I

a IFl e

a 53 S e

S v

eb. ~;*

~

0 3

-~

4:

s i

1 TINU 52 H

0 0

A 5

0 0

5 Y

2 2

1 O

U

- E 5neouC* E g >o.e EEeE:e.bc

.o Q

i i

n ES l'l'

4 INDEX DEFINITIONS i

SECTION PAGE 1.0 DEFINITIONS 1.1 ACTI0N............................................................

1-1 1.2 AXIAL FLUX DIFFERENCE.............................................

1-1 1.3 BYPASS LEAKAGE PATH...............................................

1-1 1.4 CHANNEL CALIBRATION...............................................

1-1 R63 1.5 CHANNEL CHECK.....................................................

1-1 1.6 CHANNEL FUNCTIONAL TEST...........................................

1-2 1-2 1.7 CONTAINMENT INTEGRITY... h..-

v

-v 1.8 CONTROLLEDLEAKAGE.(E).q4)(Ty[9)...................................

1-2

~

s 1.9 CORE ALTERATION...................................................

1-2 1.10 CORE OPERATING LIMIT REP 0RT.......................................

1-2 1.11 DOSE EQUIVALENT I-131............................................

1-3 1.12 E-AVERAGE DISINTEGRATION ENERGY...................................

1-3 1.13 ENGINEERED SAFETY FEATURE RESPONSE TIME...........................

1-3 1.14 FREQUENCY N0TATION................................................

1-3 1.15 GASEOUS RADWASTE TREATMENT SYSTEM.................................

1-3 1.16 IDENTIFIED LEAKAGE................................................

1-4

g14, 1.17 MEMBERS OF THE PUBLIC.............................................

1-4 1.18 0FFSITE DOSE CALCyLATION MANUAL...................................

1-4 1.19 OPERABLE - OPERABILITY............................................

1-4 1.20 OPERATIONAL MODE - M0DE...........................................

1-5 1.21 PHYSICS TESTS.....................................................

1-5 1.22 PRESSURE BOUNDARY LEAKAGE.........................................

1-5 1.23 PROCESS CONTROL PR0 GRAM...........................................

1-5 SEQUOYAH - UNIT 2 I

Amendment No.144 63 March 30, 1992 hh! ) h Th9k

1 INDEX j

LIMITING CONDITIONS FOR_0PERATION AND SURVrILLANCE RE0VIREMENTS SECTION EME 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation...............................

3/4 4-1 Hot Standby...............................................

3/4 4-2 Hot Shutdown..............................................

3/4 4-3 Cold Shutdown.............................................

3/4 4-5 i

3/4.4.2 SAFETY VALVES - SHUTD0WN..................................

3/4 4-6 l

3/4.4.3 3AFETY AND RELIEF VALVES - OPERATING

~

4 Safety Val ve s 0pe rat i ng...................................

3/4 4-7 Rel i e f Val ve s 0pe rat i ng..................................

3/4 4-8 3/4.4.4 PRESSURIZER...............................................

3/4 4-9

)

I 3/4.4.5 STEAM GENERATORS..........................................

3/4 4-10 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAG Leakage Detection 1y:M;s. b. D l d.@.i............

3/4 4-17 3/4 4-18 Op e ra t i on al L e a k age........................Er.............

ewM Cooter Sy.urM /kcuun 120tdriew furLew er..

3/YV-/

-21 3/4.4.

C M 3/4.4.8 SPECIFIC ACTIVITY.........................................

3/4 4-24 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System....................................

3/4 4-28 R138 Pressurizer................................................

3/4 4-31 RIS8 3/4 4-32

'l 3/4.4.10 DELETED 3/4.4.11 REACTOR COOLANT SYSTEM HEAD VENTS........................

3/4 4-33 R138 3/4.4.12 OVERPRESSURE PROTECTION SYSTEMS...........................

3/4 4-34 l R147 August 22, 1995 SEQUOYAH - UNIT 2 VI Amendment No. 106, 120, 138. 19T, 147 i

o i

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0UIREMENTS SECTION PAGE 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS i

Cold Leg Injection Accumulators.............

3/4 5-1

)

l

-Deleted.........................

3/4 5-3 i

3/4.5.2 ECCS SUBSYSTEMS - T., greater than or equal to 350*F 3/4 5-4 3/4.5.3 ECCS SUBSYSTEMS - T,less than 350*F 3/4 5-8 j

3/4.5.4 DELETED.........................

3/4 5-10 3 4.5.

REFUELING' WATER STORAGE TANK..........,...

3/4 5-11

~

3/y.s.4 SEAL INJh T/06 Flow.

.3/4.5-jz

/4.6 CONTAIN SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT j

Containment Integrity................... 3/4 6-1 R167 Secondary Containment Bypass Leakage 3/4 6-2 l

Containment Air Locks..................

3/4 6-7 Internal Pres sure..................... 3/4 6-9 Air Temperature...................... 3/4 6-10 Containment Vessel Structural Integrity.......... 3/4 6-11 Shield Building Structural Integrity 3/4 6-12 Emergency Gas Treatment System (Cleanup Subsystem) 3/4 6-13 j

3/4 6-15 Containment Ventilation System 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray Subsystems 3/4 6-16 R140 Lower Containment Vent Coolers 3/4 6-16b R61 l

J SEQUOYAH - UNIT 2 VII Amendment No. 59, 61, 131, 140, 167 February 10, 1994

INDEX BASES SECTION PRE 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE.............................

B 3/4 4-4 3/4.4.7 CHEMISTRY..................................................

B 3/4 4-5

.3/4.4.8 SPECIFIC ACTIVITY..........................................

B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS................................

B 3/4 4-6 3/4.4.10 STRUCTURAL INTEGRITY......................................

B 3/4 4-14 3/4.4.11 REACTOR COOLANT SYSTEM HEAD VENTS.........................

B 3/4 4-15 R147 3/4.4.12 OVERPRESSURE PROTECTION SYSTEMS...........................

B 3/4 4-15 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS...............................................

B 3/4 5-1 and 3/4.5.3 ECCS SUBSYSTEMS..........................'......

B 3/4 5-1 3/4.5.2

~

3/4.5.4 BORON INJECTION SYSTEM.....................................

B 3/4 5-2 B 3/4 5-3 ING WATER STnoan 3/4,s.h SEAL NIEC770N FLotJ. -.

6.sN S-Y

~

^

3/4.6 CONTAINMENT SYSTEMS i

)

3/4.6.1 PRIMARY CONTAINMENT........................................

B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLINGSYSTEMS.......................

B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES...............................

B 3/4 6-3 3/4.6.4 COMBUSTIBLE GAS CONTR0L....................................

B 3/4 6-fi 3/4.6.5 ICE CONDENSER..............................................

B 3/4 6-4 R188 3/4.6.6 VACUUM RELIEF LINES........................................

B 3/4 6-6 1

3/4.7 PLANT SYSTEMS B 3/4 7-1 3/4.7.1 TURBINE CYCLE..............................................

3/4.7.2 STEAM GENERATOR PRESSURE /TEMERATURE LIMITATION.............

B 3/4 7-3 3/4.7.3 COMPONENT COOLING WATER SYSTEM.............................

B 3/4 7-3 April 28, 1995 SEQUOYAH - UNIT 2 XIII Amendment No. 147, 183

.t>

4

n q

t DEFINITIONS CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be:

a.

Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

b.

Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

c.

Digital channels - the injection of a simulated signal into the chan-R132 nel as close to the sensor input to the process racks as practicable to verify OPERABILITY including alarm and/or trip functions.

CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

a.

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

1)

Capable of being closed by an OPERABLE containment automatic s

isolation valve system, or 2)

Closed by manual valves, blind flanges, or deactivated automatic g193 valves secured in their closed positions, except for valves that are open under administrative control as permitted by Specifica-tion 3.6.3.

b.

All equipment hatches are closed and sealed.

Each air lock is in compliance with the requirements of Specification l gity c.

3.6.1.3, d.

The containment leakage rates are within the limits of Specifica-R167 tion 4.6.1.1.c, The sealing mechansim associated with each penetration (e.g., welds, e.

bellows, or 0-rings) is OPERABLE, and f.

Secondary containment bypass leakage is within the limits of Speci-fication 3.6.1.2.

CONTROLLED LEAKAGE 7h5 DEWNmW M5 /ErN har7ED

.b k'W s

CORE ALTERATION R191 1.9 CORE ALTERATIONS shall be the movement of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the head removed and fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMIT REPORT 1.10 - The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document I146 R

that provides core operating limits for the current operating reload cycle.

These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.14.

Unit operation within these operating limits is addressed in individual specifications.

June 13, 1995 SEQUOYAH - UNIT 2 1-2 Amendment No. 63,117,132,146,167, 191, 193

]

DEFINITIONS IDENTIFIED LEAKAGE 1.16 IDENTIFIED LEAKAG R146 Leakage, (crept C'"' TROLLED LE^51.CE) St: :100:d y:t:::, such astpump a.

seals or valve packing 1::h: hat p eg aptured and conducted to a sump or collecting tank, or (Enre a

foauwr&w,p um

$$lwrE

$oTrYi$i, are bot _

b.

Leakag specifically located and known eltier not to interfere with the operation of leakage detection systems or not to be PRESSURE B0UNDARY LEAKAGE, or c.

' Reactor coolant system leakage through a steam generator to the secondary system.

MEMBER (S) 0F THE PUBLIC 1.17 MEMBER OF THE PUBLIC means an individual in a controlled or unrestricted area. However,- an individual is not a member of the public during any period R165 in which.the individual receives an occupational dose.

OFFSITE DOSE CALCULATION MANUAL 1.18 The OFFSITE DOSE CALCULATION MANUAL (0DCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radio-R134 I

active gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduct of the Radiologi-cal Environmental Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Pro-grams required by Section 6.8.4 and (2) descriptions of the information that R169l

)

should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.6 R159 and 6.9.1.8.

OPERABLE - OPERABILITY 1.19 A system, subsystem, train, or component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s),

and when all necessary attendant instrumentation, controls, a normal and an emergency electrical power source, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, com-ponent or device to perform its function (s) are also capable of performing their related support function (s).

March 31, 1994 SEQUOYAH - UNIT 2 1-4 Amendment No. 63, 134, 146, 159, I65, 169

I DEFINITIONS i

. SOLIDIFICATION 1.32 Deleted.

lR14i SOURCE C!lECK 1.33 Deleted.

l R14<

STAGGERED TEST BASIS 1.34 A STAGGERED TEST BASIS shall consist of:

lR14(

A test schedule for n systems, subsystems, trains or other designated a.

components obtained by dividing the specified test interval into n equal subintervals, b.

The testing of one system, subsystem, train or other designated co..,2nent at the beginning of each subinterval.

THERMAL POWER 1.35 THERMAL POWER shall be the total reactor core heat transfer rate to the h146 reactor coolant

?..

C(&cerr A6xcros Cootgpr Ibw).Sent Wntrx UNIDENTIFIED LEAKAGE zggen,g 04 tegxorp rgr s

f 1.36JIDEWIEDLEAKAGEshallbeallleakagewhichisnotIDENTIFIEDLEAKAGE, l R14e UNRESTRICTED AREA 1.37 An UNRESTRICTED AREA shall be any area, et or beyond the site boundary to h146 which access is not controlled by the licensee for purposes of prote-tion of individuals from exposure to radiation and radioactive materials or any area R63 within the site boundary used for residential quarters or industrial, commer-cial, institutiona.1, and/or recreational purposes.

SEQUOYAH - UNIT 2 1-7 Amendment No. 63, 134, MNh30,1992

REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION C'/:T IA/3rg/ga7ENTNT/CA/

A LIMITING CCNDITION FOR OPERATION ImtMunewinnw 3.4.6.1 The following Reactor Coolant System leakage detection eye +e-r shall a

be OPERABLE:

Out a.

.he-lower containment atmosphere -partice!:te-radioactivity monitor 4*g.

y: t:=- (&dsious On NATHtit.ME), ANO n

1 Out b.

T h containment pocket sump level monitor,ing cyct:m, and-

-c.

The 'cHe-centai-ment et c:phere gacecer -adioact "4ty cai te-ing 3

y:t =

j APPLICABILITY:

MODES 1, 2, 3 and 4.

ACTION /7,'

di ly wo of he bov r ui ed ea ge et ti n s ste OF RA t,fope a-on ay o in e f r u to 30 ay pr id d g ab amp es f t e c nt 4nm t

[

at sp _re ar ob* ine a a aly ed t I as on p

24 hou s w en he i

r ui ed as ous or rt'cul te ad' ac ve on or g st i

in er le th wi e,-

e at e ti T AN Y 'ith'n t e n t ho s d 'n C D S_ y 0 N ith', t f lo in 30,our.

H

tu, &gg7. j A

s SURVEILLANCE REQUIREMENTS p

~Dasr&MwrMTraN 4.4.6.1 The leakage detection a,;tma shall be demonstrated OPERABLE by:

Pencornucor a.A/helowercontainmentatmospheregaseousandparticulatemonitorkg.

.:yt t^- p r f:rm;r.:: Of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3, and Ekn nests Of b.

A ontainment pocket sump level monitorf r,; :y:t= p:-for :nce of CHANNEL ALIBRATION at least once per 18 months.

j w

s e - n -tiern-l2 flouxs Ann Esmausamwr Of I

JNor Reamtsa AraDy drnre Orannw.

\\

SEQUOYAH - UNIT 2 3/4 4-17 I

i

i Insert 1 ACTIONS:

a.

With both containment pocket sump monitors inoperable, operation may. continue for up to 30 days provided

-SR 4.4.6.2.1 is performed once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months *;

otherwise, be'in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . 'The provisions of Specification 3.0.4 are

'not' applicable.

b.

With both lower containment atmosphere radioactivity' monitors (gaseous and particulate) inoperable, operation may continue for up to 30 days provided grab samples of the lower containment atmosphere are analyzed once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or SR 4.4.6.2.1 is performed once-per.24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months *; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDONN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

The provisions of Specification 3.0.4 are not applicable.

c.

With both containment pocket sump monitors and both lower containment atmosphere radioactivity monitors inoperable, be in.at least HOT. STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

i l

c

1 REACTOR CCOLANT SYSTEM.

OPERATIONAL RAKAGE i

LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to a.

No PRESSURE BOUNDARY LEAKAGE, b.

1 GPM UNIDENTIFIED LEAKAGE, c.

150 gallons per of rimary-to-secondary leakage through any one R213 steam generatoz;, Av

~

" ^-

d.

10 GPM IDENTIFI GE fr m t stem. an+

- - ~ _ ~,-.....-. -. - - -. ~,........ w.

-2225 20 prig.

f.

1 CO." leakage at 2 Reistor C clant Syster preceure Of 2225 1 20 prig

-fr:r ny Recetcr C001:nt Syst r Prc Cur ! 012 tic V lv Op Cificd in T ble 2.t 1.

3 and 4 APPLICABILITY: MODES 1, 2,

ACTION:

a.

With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within i

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , b.

With any Reactor Coolant System leakage greater the.n any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, ::d 1cch g: f:::

n::t:rCclut cycter rr:::ur: ! 12tien Valter, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Wi y ea or Coe an Sy te Pr ssu e I ol io Va e eak ge g at r an th ab ve im't, so te the hig pr seu e rt n f e ff te s te fr m

.e ow re ur po tio wi in 4 h urs by se f t eas t o e os ua or de tiv te au ma ic alv s, r

e n t as H S

O Y w th tenxc ho rs nd n C LD

/

S it in he el owi g 3 h rs

/

A

^

f SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be d r rtrat d to be within each of the above limits by/

/extrito W f

^

R213 April 9, 1997 SEQUOYAH - UNIT 2 3/4 4-18 Amendment No. 211, 213

REACTOR COOLANT SYSTEM i

l i

SURVEILLANCE REQUIREMENTS (Continued) i Mon' ori g e c ntai ment oc ts pi ent y an dis arg at

]

1 st ce er 2h rs.

4 ea re nt ft CON ROL DL KAG to

.e re ctor cool nt p mp se s

en he act Co ant Syst m pr ssur is 35 20 ig 1 as onc per 31 d ys w th em ula ng v ve lly open Th ro si so Spe fic ion 4.0. fare ot plic le or e try to Ho e 3 or 4..

/

f

/erformanceofaReactorCoolantSystemwaterinventorybalanceat-least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .8 oni orin the eac or h ad f nge eak f s' tem t1 st a ce p j

24 our l

4 4.6

.2 ac eac or C olan Sys em P ssur Is ati Va e sp ciff in I

ab. 3.

1s 1

de nstr ted PERA E p rsua t to Spec

fica on 4

.5, ex ept at li of any aka te ing equi eme sr uire by S ecif' a-on

.0.5, each valv sha 1 be demo stra d0 RAB by erif ing aka to ew hin ts mit-a.

A lea onc per 18 m nths b.

rio to nte gM E2 hen er ep nt h s be in LO UTD0 for 72 urs r mo e an if eaka e te ting as n be per orme i

the rev us mon s.

c.

Pri to etur ing he v ive o se vice ollo ng int ance rep r or epl eme wo on he lve ithi 24 our fol win alv act tion due au mati or anuafj act no flo thr gh ev ve.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

" IN.S EM 2

Nor Asawsto n Be Pexfoxess //cra /2 //oun sen Errx8 usanavr Of Sirnoy SrHir Orexstrysw.

f

^

j SEQUOYAH - UNIT 2 3/4 4-19

Insert 2 4.4.6.2.2 Verify steam generator tube integrity is in accordance with the requirements of Technical Specification 3/4.4.5 " Steam Generators."

e 9

m Wg 4

4 b

e e

O F

)

1 I

l l

o l

-l k$[

REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVE LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.3 Leakage from each Reactor Coolant System Pressure Isolation Valve, specified in Table 3.4-1, shall be equivalent to s 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm at a Reactor

~

Coolant System pressure a 2215 psig and s 2255 psig.

APPLICABILITY: MODES 1, 2,

and 3, MODE 4, except valves in the residual heat removal system flow

. path when in, or during the transition to or from, the residual heat removal mode of operation.

ACTIONS:

a.

With one or more flow paths with leakage from one or more Reactor Coolant System Pressure Isolation Valves greater than the above limit,' isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one closed manual, deactivated automatic, or check valve

and restore the inoperable Reactor Coolant System Pressure Isolation' Valve to OPERABLE status within the following 68 hours7.87037e-4 days 0.0189 hours 1.124339e-4 weeks 2.5874e-5 months . Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , b.

Separate entry into the above ACTION is' allowed for each flow path, c.

Entry into the applicable ACTIONS for systems made inoperable by.an inoperable Reactor Coolant System Pressure Isolation Valve is required.

SURVEILLANCE REQUIREMENTS 4.4.6.3 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE pursuant to Specification 4.0.5, except that in lieu of any leakage testing requirements required by Specification 4.0.5, each valve shall be demonstrated OPERABLE by verifying

\\

leakage to be within its limit's a.

At least once per 18 months b.

Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN I

for 7 days or more and if leakage testing has not been performed in 1

the previous 9 months.

I c.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve.

'Not required to be performed in MODES 3'and 4.

Each valve used to satisfy ACTION a must have been verified to meet the

. Surveillance Requirement 4.4.6.3 and be in the reactor coolant pressure l-boundary.

Not required to be performed on Reactor Coolant System Pressure Isolation

-Valves located in the Residual Heat Removal flow path when in the shutdown cooling mode of operation.

M

^

P-SEQUOYAH - UNIT 2 3/4 4-19 Amendment No.

Nen)

PAGr EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.6 SEAL INJECTION FLOW LIMITING CONDITION FOR OPERATION 3.5.6 Reactor coolant pump seal injection flow shall be within limits.

APPLICABILITY: MODES 1, 2,

and 3.

-ACTION:

'With reactor coolant pump seal injection flow not within limit, adjust manual seal injection throttle valves to give a flow within limit in accordance with Surveillance Requirement 4.5.6 within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.6 At least once per 31 days

verify manual seal injection throttle valves are adjusted to give a flow within the emergency core cooling system safety analysis limits.

This surveillance is not required to be performed until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the reactor coolant system pressure stabilizes at = 2215 psig and s 2255 psig.

A.

A

_f SEQUOYAH - UNIT 2 3/4 5-12 Amendment No.

i 1

b/S kSE b hf.S $ 6~ F N IWDt'S6TfD b)' I//&~ $77MCH6% kCC.

REACTOR COOLANT SYSTEM BASES 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specificatio are provided to monitor and detect leakage from the Reactor Coolant Pres ure Boundary; 'These detection systems are consistent with the recomme ations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage etection Systems," May 1973.

f 3/4.4.6.2 OPERATIONAL LEAKAGE l

Industry experience has shown that while a limited am t of leakage is I

expected from the RCS, the unidentified portion of this le age can be reduced

)

to a threshold value of less than 1 GPM.

This threshold alue is sufficiently J

low to ensure early detection of additional leakage.

f The survel.llance requirements for RCS Pressure solation Valves provide

~

added assurances of valve integrity thereby reducin the probability of gross valve failure and consequent intersystem LOCA.

Le age from the RCS isolation valves is IDENTIFIED LEAKAGE and will be consider d as a portion of the allowed limit.

The 10 GPM IDENTIFIED LEAKAGE limitatio provides allowance for a limited amount of' leakage from known sources whose p esence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the 1 kage detection systems.

.The CONTROLLED LEAKAGE limitation entricts operation when the total flow supplied to the reactor coolant pump s is exceeds 40 GPM with the modulating valve in the supply line fully open a a nominal RCS pressure of 2235 psig.

This limitation ensures that in the ent of a LOCA, the safety injection flow will not be less than assumed in t accident analyses.

)

The total steam generator e leakage limit of 600 gallons per day for R213 I

all steam generators and 150 ga ons per day for any one steam generator will minimize the potential for a gnificant leakage event during steam line break.

Based on the NDE uncertainti

, bobbin coil voltage distribution and crack growth rate from the previo a inspection, the expected leak rate following a steam line rupture is lim ed to below 8.21 gpm at atmospheric conditions and 70 *F in the faulted loo, which will limit the calculated offsi';e doses to R227 within 10 percent of th 10 CFR 100 guidelines.

If the projected and cycle distribution of crack - dications results in primary-to-secondary leakage greater than 8.21.gp in the faulted loop during a postulated steam line brea event, additional t es must be removed from service in order to reduce the postulated primary o-secondary steam line break leakage to below 8.21 gpm.

The 150-g lons per day limit incorporated into SR 4.4.6 is more restrictive th the standard operating leakage limit and is intended to provide an ad stional margin to accommodate a crack which might grow at a greater tha expected rate or unexpectedly extend outside the thickness of the R227 tube suppo plate.

Hence, the reduced leakage limit, when combined with an effective eak rate monitoring program, provides additional assurance that, should a significant leak be experienced, it will be detected, and the plant shut d in a timely manner, PRESSURE BOUNDARY LEAKAGE of any magnitudo is unacceptable since it may j

be ndicative of an impending gross failure of the pressure boundary.

T refore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to promptly placed in COLD SHUTDOWN.

-/

lR213 November 17, 1990 A.

SEQUOYAH - UNIT 2 B 3/4 4-4 Amendment No. 211, 213, 227

r REACTOR COOLANT SYSTEM BASES 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION INSTRUMENTATION

' BACKGROUND GDC 30 of Appendix A to 10 CFR 50 (Ref. 1) requires means for detecting and, to the extent practical, identifying the location of the source of reactor coolant system (RCS) leakage.

Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.

Leakage detection systems must have the capability to detect significant reactor coolant pressure boundary (RCPB) degradation as soon after occurrence as practical to minimize the potential for propagation to a gross failure.

Thus, an early indication or warning signal is necessary to permit proper evaluation of all UNIDENTIFIED LEAKAGE.

Industry practice has shown that water flow changes of 0.5 to 1.0 gpm can be readily detected in contained volumes by monitoring changes in water level, in flow rate, or in the operating frequency of a pump.

The containment pocket sump used to collect UNIDENTIFIED LEAKAGE is instrumented to alarm for increases of 1.0 gpm in the normal flow rates within one hour.

This sensitivity is acceptable for detecting increases in UNIDENTIFIED LEAKAGE.

The reactor coolant contains radioactivity that, when released to the containment, can be detected by radiation monitoring instrumentation.

Reactor coolant radioactivity levels will be low during initial reactor startup and for a few weeks thereafter, until activated corrosion products have been formed and fission products appear from fuel element cladding contamination or cladding defects.

4 Instrument sensitivities of 10 pCi/cc radioactivity for 4

particulate monitoring and of 10 pCi/cc radioactivity for gaseous monitoring are practical for these leakage detection systems.

Radioactivity detection systems are included for l

monitoring both particulate and gaseous activities because of their sensitivities and rapid responses to RCS leakage.

An increase in humidity of the containment atmosphere would indicate release of water vapor to the containment.

Dew i

point temperature measurements can thus be used to monitor i

humidity levels of the containment atmosphere as an I

indicator of potential RCS leakage.

{

Since the humidity level is influenced by several factors, a quantitative evaluation of an indicated leakage rate by this j

means may be questionable and should be compared to observed i

increases in liquid flow into or from the containment sump.

Humidity level monitoring is considered most useful as an indirect alarm or indication to alert the operator to a potential problem.

Humidity monitors are not required by this LCO.

l SEQUOYAH - UNIT 2 B 3/4 4-4 Amendment No.

s

-p REACTOR COOLANT SYSTEM BASES l

Air temperature and pressure' monitoring methods may also be used to infer UNIDENTIFIED LEAKAGE to the containment.

Containment. temperature and pressure fluctuate slightly i-during plant operation, but a rise above the normally indicated range of values may indicate RCS leakage into the containment. The relevance of temperature and pressure measurements are affected by containment free volume and, for temperature, detector location. Alarm signals from these instruments can'be valuable in' recognizing rapid and yf sizable leakage to the containment.

Temperature and

' pressure monitors are not required by this LCO.

4 APPLICABLE The need to evaluate the severity of an alarm or an SAFETY ANALYSES-indication is important to the operators, and the ability to compare and verify with indications from other systems is necessary.

The, system response times and sensitivities are described in the FSAR (Ref. 3).

Multiple instrument a

locations are utilized, if needed, to ensure that the transport delay time of the leakage from its source to an instrument location yields an acceptable overall response time.

The safety significance of RCS. leakage varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring RCS leakage into the containment area is necessary. Quickly separating the IDENTIFIED LEAKAGE from the UNIDENTIFIED LEAKAGE provides quantitative information to the orsrators, allowing them to take corrective action should a leakage occur _ detrimental to the

,xclusions to the E

safety of the unit and the public.

requirements of General Design Criteria.4, for the dynamic effects of the RCS piping, have been utilized based on the leak detection capability to identify leaks before'a pipe break would occur.

RCS leakage detection instrumentation satisfies Criterion 1

~

of the NRC Policy Statement.

LCO One method of protecting against large RCS leakage derives from the ability of, instruments to rapidly detect extremely small leaks. This LCO requires instruments of diverse 1 monitoring principles to be OPERABLE to provide a high degree of confidence that extremely small leaks are detected in time to allow actions to place the plant in a safe condition, when RCS leakage indicates possible RCPB degradation.

The LCO is satisfied when monitors of diverse measurement means are available. Thus, one containment pocket sump monitor, in combination with a gaseous or particulate radioactivity monitor, provides an acceptable minimum.

SEQUOYAH - UNIT 2 B 3/4 4-4a Amendment No.

I 1

f REACTOR COOLANT SYSTEM I

BASES

.........................--......-----.---===---===.......--...............

APPLICABILITY Because of elevated RCS temperature and pressure in MODES 1, 2,

3, and 4, RCS leakage detection instrumentation is required to be OPERABLE.

In MODE 5 or 6, the temperature is to be 5 200*r and l

pressure is maintained low or at atmospheric pressure.

Since the temperatures and pressures are far lower than those for MODES 1, 2,

3, and 4, the likelihood of leakage and crack propagation are much smaller. Therefore, the requirements of this LCO are not applicable in MODES 5 and 6.

ACTIONS Action a:

With both containment pocket sump monitors' inoperable, no

~

other form of sampling can provide the equivalent information; however, the containment atmosphere radioactivity monitor will provide indications of changes in leakage. Together with the atmosphere monitor, the periodic surveillance for RCS water inventory balance, Surveillance 4.4.6.2.1, must be performed at an increased frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to provide information that is adequate to detect leakage.

A footnote is added allowing that SR 4.4.6.2.1 is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady state operation (stable pressure, temperature, power level, pressurizer and makeup tank levels, makeup, letdown, and RCP seal injection and return flows). The 12-hour allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

Restoration of the required pocket sump monitor to OPERABLE status within a Completion Time of 30 days is required to regain the function after the monitor's failure.

This time is acceptable, considering the frequency and adequacy of the RCS water inventory balance required by Action a.

Action a is modified by a note that indicates that the provisions of LCO 3.0.4 are not applicable.

As a result, a MODE change is allowed when the containment sump monitor is inoperable. This allowance is provided because other instrumentation is available to monitor RCS leakage.

l If the requirements of Action a cannot be met, the plant must be brought to a MODE in which the requirement does not i

j l

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 the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-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.

l SEQUOYAH - UNIT 2 B 3/4 4-4b Amendment No.

r REACTOR COOLANT SYSTEM BASES Action b:

With both gaseous and particulate containment atmosphere radioactivity monitoring instrumentation channels inoperable, alternative action is required.

Either grab samples of the containment atmosphere must be taken and analyzed or water inventory balances, in accordance with Surveillance 4.4.6.2.1, must be performed to provide alternate periodic information.

With a sample obtained and analyzed or water inventory balance performed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the reactor may be operated for up to 30 days to allow restoration of at least one containment atmosphere radioactivity monitor.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval provides periodic information that is adeqQate to detect leakage. A footnote is -added allowing that SR 4.4.6.2.1 is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after est'ablishing steady state operation (stable pressure, temperature, power level, pressurizer and makeup tank levels, makeup, letdown, and RCP seal injection and return flows).

The 12-hour allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

The 30 day Completion Time recognizes at least one other form of leakage detection is available.

I Action b is modified by a note that indicates that the i

{

provisions of LCO 3.0.4 are not applicable. As a result, a MODE change is allowed when the gaseous and particulate containment atmosphere radioactivity monitor channels are inoperable. This allowance is provided because other instrumentation is available to monitor for RCS leakage.

If the requirements of Action b cannot be met, the plant must be brought to a MODE in which the requirement 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 the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-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.

Action c:

With all required monitors inoperable, no automatic means of monitoring leakage are available, and immediate plant shutdown to a MODE in which the requirement does not apply is required. 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 the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SEQUOYAH - UNIT 2 B 3/4 4-4c Amendment No.

r 1

l i

REACTOR COOLANT SYSTEM j

BASES i

SURVEILLANCE Surveillance 4.4.6.1.a REQUIREMENTS This surveillance requires the performance of a CHANNEL CHECK of the required containment atmosphere radioactivity monitors. The check gives reasonable confidence that the monitors are operating properly.

The frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is based on instrument reliability and is reasonable for detecting off normal conditions.

This surveillance requires the performance of a CHANNEL CALIBRATION for the required containment atmosphere radioactivity monitors.

The calibration verifies the accuracy of the instrument string, including the instruments located inside containment.

The frequency of 18 months is a typical refueling cycle and considers channel reliability.

Operating experience has proven that this frequency is acceptable.

This surveillance requires the performance of a CHANNEL FUNCTIONAL TEST on the required containment atmosphere radioactivity monitors.

The test ensures that the monitors can perform their functions in the desired manner.

The test verifies the alarm setpoint and relative accuracy of the instrument string.

The frequency of 92 days considers instrument reliability, and operating experience has shown that it is proper for detecting degradation.

The surveillance frequencies for these tests are specified in Table 4.3-3.

Surveillance 4.4.6.1.b This surveillance requires the performance of a CHANNEL CALIBRATION for the required containment pocket sump monitors.

The calibration verifies the accuracy of the instrument string, including the instruments located inside containment.

The frequency of 18 months is a typical refueling cycle and considers channel reliability. Again, operating experience has proven that this frequency is acceptable.

REFERENCES 1.

10 CFR 50, Appendix A, Section IV, GDC 30, 2.

3.

ESAR, Sections 5.2.7 "RCBP Leakage Detection Systems" and 12.2.4 " Airborne Radioactivity Monitoring."

i f

l SEQUOYAH - UNIT 2 B 3/4 4-4d Amendment No.

i REACTOR COOLANT SYSTEM j

BASES 3/4.4.6.2 OPERATIONAL LEAKAGE BACKGROUND Components that contain or transport the coolant to or from l

the reactor core make up the reactor coolant system (RCS).

Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from j

the RCS.

During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant leakage, through either normal operational wear or mechanical deterioration.

l The purpose of the RCS Operational leakage LCO is to limit system operation in the presence of leakage from these sources to amounts that do not compromise safety.

This LCO specifies the types and' amounts of leakage.

10 CFR 50, Appendix A, GDC 30 (Ref. 1), requires means for detecting and,,to the extent practical, identifying the source of reactor coolant leakage.

Regulatory Guide 1.45 j

a (Ref. 2) describes acceptable methods for selecting leakage g

detection systems.

j The' safety significance of RCS LEAKAGE varies widely depending on its source, rate, and durat!.m..

Therefore, detecting and monitoring reactor coolant leakag= into the containment area is necessary. Quickly separating the identified LEAKAGE from the unidentified leakage is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility j

and the public.

A limited amount of leakage inside containment ir expected from auxiliary systems that cannot be made 100% leaktight.

j Leakage from these systems should be detected, located, and l

isolated from the containment atmosphere, if possible, to i

not interfere with RCS leakage detection.

l This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded.

The consequences of violating this LCO include the possibility of a los,s of coolant accident (LOCA).

l i

APPLICABLE Except for primary-to-secondary leakage, the safety analyses SAFETY ANALYSES do not address operational leakage.

However, other operational leakage is related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event. The safety analysis for an event resulting l

I' in steam discharge to the atmosphere assumes a 1 gpm primary to secondary leakage as the initial condition.

1 SEQUOYAH - UNIT 2 B 3/4 4-4e Amendment No. 211, 213, j

227, l

1

1' l

I l

l REACTOR COOLANT SYSTEM BASES Primary to secondary leakage is a factor in the dose l

l releases outside containment resulting from a steam line 1

break (SLB) accident.

To a lesser extent, other accidents or transients involve secondary steam release to the

' atmosphere, such as a steam generator. tube rupture (SGTR).

]

.The leakage contaminates the secondary fluid.

The FSAR (Ref. 3) analysis for SGTR assumes the contaminated secondary fluid is released via safety valves for up to i

30 minutes. Operator action is taken to isolate the affected steam generator within this time period. The 1.gpm primary to secondary leakage is relatively inconsequential.

The SLB is more limiting for site radiation releases. The safety analysis for the SLB accident. assumes 1 gpm primary to. secondary leakage in one generator as an initial condition. The dose consequences'resulting from the SLB i

accident are we,ll within the limits defined in 10 CFR 100 or j

the staff approved licensing basis (i.e.,

a small fraction of these limits).

Based on the NDE uncertainties, bobbin i

coil voltage distribution and crack growth rate from the previous inspection, the expected leak rate following a steam line rupture is limited to below 8.21 gpm at atmospheric conditions and 70*F in the faulted loop, which R227 i

'will limit the calculated offsite doses to within 10 percent of the 10 CFR 100 guidelines.

If the projected and cycle distribution.of crack indications results in primary-to-secondary leakage greater than 8.21 gpm in the faulted loop during a postulated steam line break event, additional tubes must be removed from service in order to reduce the postulated primary-to-secondary steam line break leakage to below 8.21 gpm.

The RCS operational leakage satisfies Criterion 2 of the NRC Policy Statement.

LCO RCS operational leakage shall be limited to:

a.

PRESSURE BOUNDARY LEAKAGE No PRESSURE BOUNDARY LEAKAGE is allowed, being indicative of material deterioration.

Leakage of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher leakage.

Violation of this LCO could result in continued degradation of the RCPB.

Leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

j b.

UNIDENTIFIED LEAKAGE One gpm of UNIDENTIFIED LEAKAGE is allowed as a reasonable minimum detectable amount that the I

containment air monitoring and containment pocket SEQUOYAH - UNIT 2 B 3/4 4-4f Amendment No. 211, 213,

227, I

l' i

3 l

l t

REACTOR COOLANT SYSTEM BASES sump level monitoring equipment can detect within a reasonable time period.

Violation of this LCO could result in continued degradation of the RCPB, if the leakage is from the pressure boundary.

I c.

Primary to Secondary Leakage through Any One Steam Generator (SG)

The 150 gallons per day limit on one SG is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line rupture.

If leaked through many cracks, the cracks are very small, and the above assumption is conservative.

The 150-gallons per day limit incorporated into Surveillance 4.4.6.2.1 is more restrictive than the standard operating leakage limit and is intended to provide a'n additional margin to accommodate a crack a

which might. grow at a greater than expected rate or R227 unexpectedly extend outside the thickness of the tube f

support plate.

Hence, the reduced leakage limit, when combined with an effective leak rate monitoring i

program, provides additional assurance that, should a significant leak be experienced, it will be detected, and the plant shut down in a timely manner.

I d.

IDENTIFIED LEAKAGE Up to 10 gpm of IDENTIFIED LEAKAGE is considered allowable because leakage is from known sources that do not interfere with detection of UNIDENTIFIED.

LEAKAGE and is well within the capability of the RCS l

l Makeup System.

IDENTIFIED LEAKAGE includes leakage to I

the containment from specifically known and located l

sources, but does not include PRESSURE BOUNDARY LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered leakage).

l Violation of this LCO could result in continued j

degradation of a component or system.

1 k

I l

1 APPLICABILITY In MODES 1, 2, 3, an'd 4, the potential for reactor coolant PRESSURE BOUNDARY LEAKAGE is greatest when the RCS is pressurized, i

In MODES 5 and 6, leakage limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for leakage.

l 1

i f

SEQUOYAH - UNIT 2 B 3/4 4-4g Amendment No. 211, 213,

)

227, i

l l

f REACTOR COOLANT SYSTEM BASES

...........................--............--.....----........... -===.....--

LCO 3/4.4.6.3, "RCS Pressure Isolation Valve (PIV) Leakage,"

measures leakage through each individual PIV and can impact this LCO.

Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS leakage when the other is leak tight.

If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable IDENTIFIED LEAKAGE.

ACTIONS Action a:

If any PRESSURE BOUNDARY LEAKAGE exists, the reactor must be brought to lower pressure conditions to reduce the severity of the leakage and its potential consequences.

It should be noted that leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE. The reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .and MODE 5 within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

This action reduces the leakage and also reduces the factors that tend to degrade the pressure boundary.

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.

In MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.

Action b:

UNIDENTIFIED LEAKAGE, IDENTIFIED LEAKAGE, or primary-to-secondary leakage in excess of the LCO limits must be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

This completion time allows time to verify leakage rates and either identify UNIDENTIFIED LEAKAGE or reduce leakage to within limits before the reactor must be shut down.

This action is necessary to prevent further deterioration of the RCPB.

If UNIDENTIFIED LEAKAGE, IDENTIFIED LEAKAGE, or primary to secondary leakage cannot be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the reactor must be brought to lower pressure conditions to reduce the severity of the leakage and its potential consequences.

The reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

This action reduces,the leakage and also reduces the factors that tend to degrade the pressure boundary.

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.

In MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.

SEQUOYAH - UNIT 2 B 3/4 4-4h Amendment No. 211, 213,

227,

N l

REACTOR COOLANT SYSTEM BASES

....................--..............-==...----------...........----------.

SURVEILLANCE Surveillance 4.4.6.2.1 REQUIREMENTS Verifying RCS leakage to be within the LCO limits ensures the integrity of the RCPB is maintained.

PRESSURE BOUNDARY LEAKAGE would at first appear as UNIDENTIFIED LEAKAGE and can only be positively identified by inspection.

It should be noted that leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

UNIDENTIFIED LEAKAGE and IDENTIFIED LEAKAGE are determined by performance of an RCS water inventory balance.

Primary-to-secondary leakage is also measured by performance of an RCS water inventory balance in conjunction with effluent monitoring within the secondary steam and feedwater systems.

The RCS water inventory balance must be met with the reactor at steady state operating conditions (stable pressure, temperature, power level, pressurizer.and makeup tank levels, makeup, letdown, and RCP seal injection and return flows).

Therefore, a footnote is added allowing that this SR is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady state operation.

The 12-hour allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

Performance of this surveillance within the 12-hour allowance is required to maintain compliance with the provisions of Specification 4.0.3.

Steady state operation is required to perform a proper inventory balance since calculations during maneuvering are not useful.

For RCS operational leakage determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

An early warning of PRESSURE BOUNDARY LEAKAGE or UNIDENTIFIED LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment pocket sump level.

It should be noted that leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

These leakage detection systems are specified in LCO 3/4.4.6.1, " Leakage Detection Instrumentation."

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months frequency is a reae--able interval to trend leakage and recogniz,es the impc2. Ce of early leakage detection in the prevention of a. mdents.

Surveillance 4.4.6.2.2 This surveillance provides the means necessary to determine SG OPERABILITY in an operational MODE.

The requirement to demonstrate SG tube integrity in accordance with the Steam Generator Tube Surveillance Program emphasizes the importance of SG tube integrity, even though this surveillance cannot be performed at normal operating conditions.

SEQUOYAH - UNIT 2 B 3/4 4-41 Amendment No. 211, 213,

227, i

a

r

.c REACTOR COOLANT SYSTEM BASES REFERENCES 1.

2.

Regulatory Guide 1.45, May 1973.

3.

FSAR, Section 15.4.3.

SEQUOYAH - UNIT 2 B 3/4 4-4j Amendment No. 211, 213,

227,

y REACTOR COOLANT SYSTEM BASES 3/4.4.6.3 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVE LEAKAGE BACKGROUND 10 CFR 50.2, 10 CFR 50.55a(c), and GDC 55 of 10 CFR 50, Apperidix A (Refs.

1, 2, and 3), define reactor coolant system (RCS) pressure isolation valves (PIVs) as any two normally closed valves in series within the reactor coolant presse e boundary (RCPB),

which separate the high pressure RCS from an attached low pressure system.

During their lives, these valves can produce varying amounts of reactor coolant leakage through either normal operational wear or mechanical deterioration. The RCS PIV leakage LCO allows RCS high pressure operation when leakage through these valves exists in amounts that do not compromise safety.

Although this specification provides a limit on allowable PIV leakage rate, its main pufpose is to prevent overpressure failure of the low pressure portions of connecting systems.

The leakage limit is an indication that the PIVs between the RCS and the connecting systons are degraded or degrading.

PIV leakage could lead to overpressure of the low pressure piping or components.

Failure consequences could be a loss of coolant accident (LOCA) outside of containment or an unanalyzed accident that could degrade the ability for low pressure injection.

The basis for this LCO is the 1975 NRC, " Reactor Safety Study," (Ref. 4) that identified potential intersystem LOCAs as a significant contributor to the risk of core melt.

A subsequent study (Ref. 5) evaluated various PIV configurations to determine the probability of intersystem LOCAs.

PIVs are provided to isolate the RCS from the following typically connected systems:

a.

Residual Heat Removal (RHR) System; b.

Safety Injection System; and c.

Chemical and Volume Control System.

}

The PIVs are listed in Tabl*a 3. 4-1.

Violation of this LCO could result in continued degradation of a PIV, which could lead to overpressurization of a low pressure system and the loss of the integrity of a fission product barrier.

SEQUOYAH - UNIT 2 B 3/4 4-4k Amendment No.

l

1

. /

REACTOR CCOLANT SYSTEM BASES APPLICABLE Reference 4 identified potential intersystem LOCAs as SAFETY ANALYSES a significant contributo-to the risk of core melt.

The dominant accident tequence in the intersystem LOCA category is the fallere of the low pressure portion of the RHR System outside of containment.

The accident is the result of a postulated failure of the PIVs, which are part of the RCPB, and the subsequent pressurization of the RHR System downstream of the PIVs from the RCS.

Because the low pressure portion of the RHR System is typically designed for 600 psig, overpressurization failure of the RHR low pressure line would result in a LOCA outside containment and subsequent rir' of core melt.

Reference 5 evaluated various PIV configurations, leakage testing of the valves, and operational changes to determine the effect on the probability of intersystem LOCAs. Tais study concluded that periodic leakage testing of the PIVs can substantially reduce the probability of an intersystem LOCA.

RCS PIV leakage satisfies Criterion 2 of the NRC Policy Statement.

LCO RCS PIV leakage is IDENTIFIED LEAKAGE into closed systems connected to the RCS.

Isolation valve leakage is usually on the order of drops per minute.

Leakage that increases significantly suggests that something is operationally wrong and corrective action must be taken.

The LCO PIV leakage limit is 0.5 gpm per nominal inch of valve size with a maximum limit of 5 gpm.

The previous criterion of 1 gpm for all valve sizes imposed an unjustified penalty on the larger valves without providing information on potential valve degradation and resulted in higher personnel radiation exposures. A study concluded a leakage rate limit based on valve size was superior to a single allowable value.

Reference 6 permits leakage testing at a lower pressure differential than between the specified maximum RCS pressure and the normal pressure of the connected system during RCS operation (the maximum pressure differential) in those types of valves in which the higher service pressure will tend to diminish the overall leakage channel opening.

In such cases, the observed rate may be adjusted to the maximum pressure differential by assuming leakage is directly proportional to the pressure differential to the one half power.

SEQUOYAH - UNIT 2 B 3/4 4-41 Amendment No.

y REACTOR COOLANT SYSTEM BASES APPLICABILITY In MODES 1, 2,

3, and 4, this LCO applies because the PIV leakage potential is greatest when the RCS is pressurized.

In MODE 4, valves in the RHR flow path are not required to meet the requirements of this LCO when in, or during the transition to or from, the RHR mode of operation.

In MODES 5 and 6, leakage limits are not provided because the lower reactor coolant pressure results in a reduced potential for leakage and for a LOCA outside the containment.

ACTIONS Action a:

The flow path must be isolated.

Action a is modified by a note that the valves used for isoldtion must meet the same leakage requirements as the PIVs and must be within the RCPB.

Action a requires that the isolation with one valve must be performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Four hours provides time to reduce leakage in excess of the allowable limit and to isolate the affected system if leakage cannot be reduced.

The 4-hour completion time allows the actions and restricts the operation with leaking isolation valves.

The 72-hour completion time after exceeding the limic allows for the restoration of the leaking PIV to OPERABLE status.

This timeframe considers the time required to complete this action and the low probability of a second valve failing during this

period, i

If leakage cannot be reduced or the system isolated, the plant must be brought to a MODE in which the requirement does not apply.

To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

This Action may reduce the leakage and also reduces the potential for a LOCA outside the containment.

The allowed Completion Times are reasonable based on operatin experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

Action b; Action b provides clarification that each flow path allows separate entry into Action a This is allowed based upon the functional independence of the flow path.

SEQUOYAH - UNIT 2 B 3/4 4-4m Amendm

't No.

~

1 i

.r REACTOR COOLANT SYSTEM BASES-Action c:

1 Action c requires an evaluation of affected systems if a PIV is inoperable.

The leakage may have affected system operability er isolation of a leaking flow path j

with an alternate valve may have degraded the ability of the interconnected system to perform its safety L

function.

SURVEILLANCE Surveillance 4.4.6.3 REQUIREMENTS Performanca of leakage testing on each RCS PIV or isolation valve used to satisfy Action a is required to verify that leakage is below the specified limit and to identify each leaking valve.

The leakage limit o,f 0.5 gpm per inch of nominal valve diameter up to 5 gpm maximum applies to each valve.

Leakage testing requires a s. table pressure condition.

For the two PIVs'in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves.

If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the other valve in series meets the leakage requirement.

In this situation, the protection provided by redundant valves would be lost.

4 Testing ~is to be performed every 18 months, a typical refueling cycle, if the plant does not go into MODE 5 l

for at least 7 days.

The 18 month frequency is consistent with 10 CFR 50.55a(g) (Ref. 7) as contained in the Inservice Testing Program, is within frequency allowed by the American Society of Mechanical Engineers (ASME) Code,Section XI (Ref. 6), and is based on the need to perform such surveillances under the conditions that apply during an outage and the

. potential for an unplanned transient if the surveillances were performed with the reactor at power.

In addition, testing must be performed once after the

' valve has been opened by flow or exercised to ensure tight reseating..PIVs disturbed in the performance of this surveillance should also be test.ed unless

. documentation shows that an infinite testing loop cannot practically be avoided.

Testing must be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the valve has been reseated.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a reasonable and practical time limit for performing this test after opening or reseating a valve.

SEQUOYAH - UNIT 2 B 3/4 4-4n Amendment No.

i

. J REACTOR COOLANT YSTEM BASES The leakage limit is to be met at the RCS pressure associated with MODES 1 and 2.

This permits leakage testing at high differential pressures with stable conditions not possible in the MODES with lower pressures.

Entry into MODES 3 and 4 is allowed to establish the necessary differential pressures and stable conditions to allow for performance of this surveillance.

The note that allows this provision is complementary to the frequency of prior to entry into MODE 2 whenever the unit has been.in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months.

In addition, this surveillance is not required to be performed on the RHR System when the RHR System is aligned to the RCS in the shutdown cooling mode of operation.

PIVs contained in the RHR shutdown cooling flow path must be leakage rate tested after RRR is secured and stable unit conditions and the necessar'y differential pressures are established.

a REFERENCES 1.

2, 10 CFR 50.55a(c).

3.

10 CER 50, Appendix A,Section V, GDC 55.

4.

WASH-1400 (NUREG-75/014), Appendix V, October 1975.

5.

NUREG-0677, May 1980.

6.

ASME, Boiler and Pressure Vessel Code, 1

Section XI.

7.

EMERGENCY CORE COOLING SYSTEM l

l 1

1 SEQUOYAH - UNIT 2 B 3/4 4-4o Amendment No.

j

r I

,s l

BASES i

3/4.5.6 SEAL INJECTION FLOW BACKGROUND The function of the seal injection throttle valves during an accident is similar to the function of the ECCS throttle valves in that each restricts flow from the centrifugal charging pump header to the Reactor Coolant System (RCS).

The restriction on reactor coolant pump (RCP) seal injection flow limits the amount of ECCS flow that i

would be diverted from the injection path following an accident.

This limit is based on safety analysis assumptions that are required because RCP seal injection flow is not isolated during safety injection.

APPLICABLE All ECCS subsystems are taken credit for in the large SAFETY ANALYSES break loss of coolant accident (LOCA) at full power j

(Ref. 1).

The LOCA analysis establishes the minimum flow for the ECCS pumps.

The centrifugal charging l

pumps are also credited in the small break LOCA analysis. This analysis establishes the flow and discharge head at the design point for the centrifugal charging pumps. The steam generator tube repture and main steam line break event analyses also credit the centrifugal charging pumps, but are not limiting in their design. Reference to these analyses is made in assessing changes to the Seal Injection System for i

evaluation of their effects in relation to the acceptance limits-in these analyses.

This LCO ensures that seal injection flow will be sufficient for RCP seal integrity but limited so that the ECCS trains will be capable of delivering sufficient water to match bolloff rates soon enough to minimize uncovering of the core following a large j

LOCA.

It also ensures that the centrifugal charging i

pumps will deliver sufficient water fcr a small LOCA and sufficient boron to maintain the core suberitical.

For smaller LOCAs, the charging pumps alone deliver i

sufficient fluid to overcome the loss and maintain RCS inventory.

Seal injection flow satisfies Criterion 2

)

of the NRC Policy. Statement.

1 SEQUOYAH - UNIT 2 B 3/4 5-4 Amendment No.

.j l

EMERGENCY CORE COOLING SYSTEM l

BASES LCO The intent of the LCO limit on seal injection flow is to make sure that flow through the RCP seal water injection line is low enough to ensure that sufficient centrifugal charging pump injection flow is directed to the RCS via the injection points (Ref. 2).

The LCO is not strictly a flow limit, but rather a flow limit based on a flow line resistance.

In order to establish the proper flow line resistance, a pressure and flow must be known.

The flow line resistance is established by adjusting the RCP seal injection needle valves to provide a total seal I

injection flow in the acceptable region of Figure B 3.5.6-1.

The centrifugal charging pump discharge header pressure remains essentially constant through all the applicable MODES of this LCO.

A reduction in RCS pressure would result in more flow being diverted to the RCP seal injectiod line than at normal operating pressure. The valve settings established at the prescribed centrifugal charging pump discharge header pressure result in a conservative valve position should RCS pressure decrease. The flow limits established by Figure B 3.5.6-1 are consistent with the accident analysis.

The limits on seal injection flow must be met to render the ECCS OPERABLE.

If these conditions are not met, the ECCS flow will not be as assumed in the accident analyses.

APPLICABILITY In MODES 1, 2, and 3, the seal injection flow limit is dictated by ECCS flow requirements, which are specified for MODES 1, 2,

3, and 4.

The seal injection flow limit is not applicable for MODE 4 and lower, however, because l'.'gh seal injection flow is less critical as a result of the lower initial RCS pressure and decay heat removal requirements in these MODES.

Therefore, RCP seal injection flow must be limited in MODES 1, 2, and 3 to ensure adequate ECCS performance.

SEQUOYAH - UNIT 2 B 3/4 5-5 Amendment No.

r

.J EMERGENCY CORE COOLING SYSTEM l

BASES ACTION With the seal injection flow exceeding its limit, the amount of charging flow available to the RCS may be reduced.

Under this condition, action must be taken to restore the flow to below its limit. The operator i

has 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from the time the flou is known to be above the limit to correctly position the manual valves and thus be in compliance with the accident analysis. The completion time minimizes the potential exposure of the plant to a LOCA with insufficient injection flow and provides a reasonable time to restore seal injection flow within limits. This time is conservative with respect to the completion times of other ECCS LCOs; it is based on operating experience and is sufficient for taking corrective actions by operations personnel.

When the actions cannot be completed within the required completion time, a controlled shutdown must be initiated. The completion time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for reaching MODE 3 from MODE 1 is a reasonable time for a controlled shutdown, based on operating experience and normal cooldown rates, and does not challenge plant safety systems or operators.

Continuing the plant shutdown from MODE 3, an additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is a reasonable time, based on operating experience and normal cooldown rates, to reach MODE 4, where this LCO is no longer applicable.

SURVEILLANCE Surveillance 4.5.6 REQUIREMENTS Verification every 31 days that the manual seal injection throttle valves are adjusted to give a flow within the limit ensures that proper manual seal.

injection throttle valve position, and hence, proper seal injection flow, is maintained. The differential pressure that is above the reference minimum value is established between the charging header (PT 62-92) and the RCS, and total seal injection flow is verified to be within the limits determined in accordance with the ECCS safety analysis (Ref. 3).

The seal water inje,ction flow limits are shown in Figure B 3.5.6-1.

The frequency of 31 days is based on engineering judgment and is consistent with other ECCS valve surveillance frequencies.

The frequency has proven to be acceptable through operating experience.

The requirements for charging flow vary widely according to plant status and configuration. When charging flow is adjusted, the positions of the air-operated valves, which control charging flow, are SEQUOYAH - UNIT 2 B 3/4 5-6 Amendment No.

1 l

l t

.c l

EMERGENCY CORE COOLING SYSTEM 1

BASES 1

l adjusted to balance the flows through the charging header and through the seal injection header to ensure that the seal injection flow to the RCPs is maintained i

between 8 and 13 gpm per pump.

The reference minimum l

l differential pressure across the seal injection needle valves ensures that regardless of the varied settings of the charging flow control valves that are required to support optimum charging flow, a reference test condition can be established to ensure trat flows across the needle valves are within the safety analysis.

The values in the safety analysis for this reference set of conditions are calculated based on conditions during power operation and they are correlated to the minimum ECCS flow to be maintained under the most limiting accident conditions.

As noted, the surveillance is not required to be performed until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the RCS pfessure has stabilized within a i 20 psig range of normal operating pressure.

The RCS pressure requirement is specified since this configuration will produce the required pressure conditions necessary to assure that I

the manual valvr3 are set correctly.

The exception is limited to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to ensure that the surveillance is timely.

Performance of this surveillance within the 4-hour allowance is required to maintain compliance with the provisions of Specification 4.0.3.

REFERENCES 1.

FSAR, Chapter 6.3 " Emergency Core Cooling System" and Chapter 15.0 " Accident Analysis".

2.

3.

Westinghouse Electric Company Calculation CN-FSE-99-48 l

1 1

SEQUOYAH - UNIT 2 B 3/4 5-7 Amendment No.

l l

oN tne m

d 0

n 5

e mA 5

4 W

n

+'

+

s t

0

)

i 1 m 4 mp

- i

6. L g

5. ow E

(

w

.,. L o

3l

+BN.

8 l

~. A O F

F 5:

5 B

TI n

.v P G.

n o

4 Eo EE.

o

/

i 3

t i

g c

B Rt

+CR e

c C

~

j Ue A

n I

' /

l j

Gn I

a i

e Fl 5

e 3 S a

S

,r-i

^

.t

,S; y;.

03 2

T INU 52 H

0 0

0 A

5 0

5 Y

2 2

1 1

OU eGE5 mea tc* kE,_ g c8,$eE.5Ee.

c

.o Q

s n

ES

ENCLOSURE 3 TENNESSEE VALLEY AUTHORITY f

SEQUOYAH NUCLEAR PLANT (SQN)

UNITS 1 AND 2 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE 98-10 REVISED PAGES I.

AFFECTED PAGE LIST Unit 1 Unit 2 j

Index Page I Index Page I

~

Index Page VI Index Page VI Index Page VII Index Page VII Index Page XIII Index Page XIII 1-2 1-2 1-3 1-4 1-7 1-7 3/4 4-13 3/4 4-17 3/4 4-14 3/4 4-18 3/4 4-15 3/4 4-19 3/4 5-12 3/4 5-12 B 3/4 4-4a B 3/4 4-4 B 3/4 4-4b B 3/4 4-4a B 3/4 4-4c B 3/4 4-4b B 3/4 4-4d B 3/4 4-4c B 3/4 4-4e B 3/4 4-4d B 3/4 4-4f B 3/4 4-4e B 3/4 4-4g B 3/4 4-4f B 3/4 4-4h B 3/4 4-4g B 3/4 4-41 B 3/4 4-4h i

B 3/4 4-4j B 3/4 4-4i B 3/4 4-4k B 3/4 4-4j B 3/4 4-41 B 3/4 4-4k B 3/4 4-4m B 3/4 4-41 B 3/4 4-4n B 3/4 4-4m B 3/4 4-4o B 3/4 4-4n B 3/4 4-4p B 3/4 4-4o B 3/4 4-4q B 3/4 5-4 B 3/4 5-4 B 3/4 5-5 B 3/4 5-5 B 3/4 5-6 B 3/4 5-6 B 3/4 5-7 B 3/4 5-7 s-8 B 3/4 5-8 II.

v&

,D PAGES g

St

.ttached.

E3-1

1

.' t INDEK DEFINITIONS SECTION PAGE I

.1.0 DEFINITIONS i

1.1 ACTION 1-1 1.2 AXIAL FLUX DIFFERENCE 1-1

.1.3 BYPASS LEAKAGE PATH.

1-1 1.4 CHANNEL CALIBRATION.

1-1 R75

]

i 1.5 CHANNEL CHECK.

1-1

-1.6 CHANNEL FUNCTIONAL TEST 1-2 l

l 1.7 CONTAINMENT INTEGRITY.

1-2 1.8 CONTROLLED LEAKAGE (Deleted) 1-2 l

.1.9 CORE ALTERATION.

1-2 R75 1.10 CORE OPERATING LIMITS REPORT 1-2 1.11 DOSE EQUIVALENT I-131.

1-3 1

1.12

_E-AVERAGE DISINTEGRATION ENERGY.

1-3 R159

1.13 ENGINEERED SAFETY FEATURE RESPONSE TIME 1-3 1.14 FREQUENCY NOTATION 1-3 1.15 GASEOUS RADWASTE TREATMENT SYSTEM..................

1-3

-1.16-IDENTIFIED LEAKAGE 1-3 1.17 MEMBERS OF THE PUBLIC.

1-4 1.18 OFFSITE DOSE CALCULATION MANUAL 1-4 1.19 OPERABLE - OPERABILITY 1-4 1.20 OPERAT ONAL MODE - MODE 1-4 1.21 PHYSICS TESTS'.

1-4 1.22 -PRESSURE BOUNDARY LEAKAGE 1-5 1.23 PROCESS CONTROL PROGRAM 1-5 SEQUOYAH - UNIT 1 I

Amendment No. 71, 155,

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS I

SECTION PAGE

'3/4.4 REACTOR COOLANT SYSTEM' q

3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION 3/4 4-1 STARTUP AND POWER OPERATION l

HOT STANDBY.

3/4 4-la l

3/4 4-2 SHUTDOWN 3/4.4.2 SAFETY VALVES - SHUTDOWN 3/4 4-3 3/4.4.3 SAFETY AND RELIEF VALVES - OPERATING q

SAFE.TY VALVES - OPERATING 3/4 4-4 3/4 4-4a RELIEF VALVES - OPERATING 3/4 4-5 3/4.4.4 PRESSURIZER 3/4.4.5 STEAM GENERATORS 3/4 4-6 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION INSTRUMENTATION 3/4 4-13 l

3/4 4-14 OPERATIONAL LEAKAGE REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVE LEAKAGE 3/4 4-15a l

3/4.4.7' CHEMISTRY.

3/4 4-16 3/4.4.8 SPECIFIC ACTIVITY.

.3/4 4-19 3/4.4.9 PRESSURE / TEMPERATURE LIMITS

.3/4 4-23 REACTOR COOLANT SYSTEM.

. PRESSURIZER.

.3/4 4-26

-3/4.4.10 DELETED 3/4 4-27 lR212

~

.3/4 4-28 3/4.4.11 REACTOR COOLANT SYSTEM HEAD VENTS 3/4.4.12 OVER PRESSURE PROTECTION SYSTEM

.3/4 4-29 R161

\\

l i

1 SEQUOYAH - UNIT 1 VI Amendment No. 116, 133, 157, 208, 1

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

- Sr.;CTION PA_Gg Aq 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1-ACCUMULATOKS COLD LEG INJECTION ACCUMULATORS 3/4 5-1 DELETED.

3/4 5-3 3/4.5.2 ECCS SUBSYSTEMS -' T,, Greater Than or Eqtal to 350*F 3/4 5-4

~3/4.5.3 ECCS SUBSYSTEMS - T.,, Less Than 350*F 3/4 5-8 R144

.3/4.5.4 DELETED 3/4 5-10 3/4.5.5 REFUELING WATER STORAGE TANK 3/4 5-11 3/4.5.6 SEAL INJECTION FLOW.

3/4 5-12 l

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1'

-PRIMARY CONTAINMENT CONTAINMENT INTEGRITY.

3/4 6-1 SECONDARY CONTAINMENT BYPASS LEAKAGE 3/4 6-2 lR180 CONTAINMENT AIR LOCKS 3/4 6-7 l

l~

INTERNAL PRESSURE.

3/4 6-9 AIR TEMPERATURE 3/4 6-10 1.

l

' CONTAINMENT VESSEL STRUCTURAL INTEGR1TY 3/4 6-11 3/4 6-12 SHIELD BUILDING STRUCTURAL INTEGRITY EMERGENCY GAS TREATMENT SYSTEM (CLEANUP SUBSYSTEM) 3/4 6-13 CONTAINMENT VENTILATION SYSTEM 3/4 6-15

-3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SUBSYSTEMS 3/4 6-16 R154 LOWER CONTAINMENT VENT COOLERS 3/4 6-16b SEQUOYAH - UNIT 1 VII Amendment No. 67, 69, 116, 140, 150, 176,

INDEX BASES SECTION PAGE 3/4.4.6-REACTOR COOLANT SYSTEM LEAKAGE B 3/4 4-4a l

)

1 3/4.4.7 CHEMISTRY.

B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY

. B 3/4 4-5 4

3/4.4.9 PRESSURE / TEMPERATURE LIMITS B 3/4 4-6 3/4.4.10 STRUCTURAL INTEGRITY B 3/4 4-14 3/4.4.11 REACTOR COOLANT SYSTEM HEAD VENTS B 3/4 4-14 R161

]

3/4.4.12 OVERPRESSURE PROTECTION SYSTEMS B 3/4 4-14 1

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS

. B 3/4 5-1

-3/4.5.4 BORON INJECTION SYSTEM B 3/4 5-2 3/4.5.5 REFUELING WATER STORAGE TANK B 3/4 5-3

]

3/4.5.6 SEAL INJECTION FLOW.

B 3/4 5-4 l

3/4.6 CONTAINMENT SYSTEMS

'3/4.6.1 PRIMARY CONTAINMENT.

B 3/4 6-1 3/4.6.2'

'DEPRESSURIZATION AND COOLING SYSTEMS B 3/4 6-3 B 3/4 6-3

-3/4.6.3

' CONTAINMENT ISOLATION VALVES 3/4.6.4 COMBUSTIBLE GAS CONTROL.

B 3/4 6-3 3/4.6.5 ICE CONDENSER.

. B 3/4 6-4 3/4.6.6 VACUUM RELIEF LINES.

B 3/4 6-6 lR201 3/4.7 PLANT SYSTEMS 3/4.7.1-TURBINE CYCLE B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.

B 3/4 7-3 B 3/4 7-3 3/4.7.3 COMPONENT COOLING WATER SYSTEM SEQUOYAH - UNIT 1 XIII Amendment No. 157, 197,

DEFINITIONS CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be:

a.

Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

b.

Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

R145 c.

Digital channels - the injection of a simulated signal into the channel as close to the sensor input to the process racks as practicable to verify OPERABILITY including alarm and/or trip functions.

CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

a.

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

1)

Capable of being closed by an OPERABLE containment automatic isolation valve system, or 2)

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are open under administrative control as R207 permitted by Specification 3.6.3.

b.

All equipment hatches are closed and sealed, c.

Each air lock is in compliance with the requirements of R180 j

Specification 3.6.1.3, d.

The containment leakage rates are within the limits of Specification 4.6.1.1.c, e.

The sealing mechansim associated with each penetration (e.g.,

welds, bellows, or O-rings) is OPERABLE, and f.

Secondary containment bypass leakage is within the limits of Specification 3.6.1.2.

CONTROLLED LEAKAGE 1.8 This definition has been deleted.

l CORE ALTERATION 1.9 CORE ALTERATION shall be the movement of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor R205 vessel with the head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMIT REPORT 1.10 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle.

R159 These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.14.

Unit operation within these operating limits is addressed in individual specifications.

SEQUOYAH - UNIT 1 1-2 Amendment No. 12, 71, 130, 141, 155 176, 201, 203,

DOSE EOUIVALENT I-131 1.11 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie /

lR159 gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

l E - AVERAGE DISIbTEGRATION ENERGY l

1.12 E shall be the average (weighted in proportion to the concentration of lR159 each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval lR159 from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e.,

the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FREOUENCY NOTATION 1.14 The FREQUENCY NOTATION specified for the performance of Surveillance lR159 Requirements shall correspond to the intervals defined in Table 1.2.

GASEOUS RADWASTE TREATMENT SYSTEM 1.15 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed lR159 to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.16 IDENTIFIED LEAKAGE shall be:

lR159 a.

Leakage such as that from pump seals or valve packing (except reactor coolant pump seal water injection or leakoff) that is captured and conducted to collection systems or a sump or collecting tank, or SEQUOYAH - UNIT 1 1-3 Amendment No. 12, 71, 155,

UNIDENTIFIED LEAKAGE 1.36 UNIDENTIFIED LEAKAGE shall be all leakage (except reactor coolant pump seal water injection or leakoff) that is not IDENTIFIED LEAKAGE.

UNRESTRICTED AREA 1.37 An UNRESTRICTED AREA shall be any area, at or beyond the site boundary to lR159 which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or any area R75 within the site boundary used for residential quarters or industrial, commerical, institutional, and/or recreational purposes, l

VENTILATION EXHAUST TREATMENT SYSTEM 1.38 A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and lR159 installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or REPA filters for the purpose of removing iodines or parti-culates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents).

Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING 1.39 VENTING is the controlled process of discharging air or gas from a lR159 confinement to maintain temperature, pressure, humidity, concentration or other j

operating condition, in such a manner that replacement air or gas is not J

provided or required during VENTING.

Vent, used in system names, does not imply a VENTING process.

i SEQUOYAH 4

UNIT 1 1-7 Amendment No. 12, 71, 155,

l REACTOR COOLANT SYSTEM-3/4.4.6 REACTOR COOLANT SYSTEM LEANAGE LEAKAGE DETECTION INSTRUMENTATION l

JLIMITING CONDITION FOR OPERATION L

3.4.6.1 The following Reactor Coolant System leakage detection instrumentation l shall be OPERABLE:

L a.

One lower containment atmosphere radioactivity monitor (gaseous or particulate), and b.'

one containment pocket sump level monitor.

APPLICABILITY: MODES 1, 2,

3 and 4.

ACTIONS:

l With both~ containment pocket sump monitors inoperable, operation may a.

l continue for up to 30 days provided SR 4.4.6.2.1 is performed once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months *; otherwise, be in at least HOT STANDBY within the next

-6 hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The provisions of Specification 3.0.4 are not applicable.

b.

With both lower containment atmosphere radioactivity monitors (gaseous and particulate) inoperable, operation may continue for up to 30 days provided grab samples of the lower containment atmosphere are analyzed once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or SR 4.4.6.2.1 is performed once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months *; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

The provisions of Specification 3.0.4 are not applicable.

c.

With both containment pocket sump monitors and both lower containment atmosphere radioactivity monitors inoperable, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the 3

following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

]

SURVEILLANCE REQUIREMENTS 4.4.6 1 The leakage detection instrumentation shall be demonstrated OPERABLE l

by:

a.

Performance of the lower containment atmosphere gaseous and particulate monitor CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3, and b.

Performance of containment pocket sump level monitor CHANNEL l

CALIBRATION at least once per 18 months.

Surveillance performance not required until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

SEQUOYAH - UNIT 1 3/4 4-13 Amendment No. 12

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDlTION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a.

No PRESSURE BOUNDARY LEAKAGE, b.-

1 GPM UNIDENTIFIED LEAKAGE, c.

150 gallons per day of primary-to-secondary leakage through any one lR226 steam generator, and

'd.

10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System.

APPLICABILITY: MODES 1,-2, 3 and 4 ACTION:

a.

With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

1 b.

With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage l

rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY f

within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

1 SURVEILLANCE REQUIREMENTS I

4.4.6.2.1 Reactor Coolant System leakages shall be verified to be within each of the above limits by performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .*

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 R16 or 4.

4.4.6.2.2 Verify steam generator tube integrity is in accordance with the requirements of Technical Specification 3/4.4.5, " Steam Generators."

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

SEQUOYAH - UNIT 1 3/4 4-14 Amendment No. 12, 214, 222,

\\

C' t

HEACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVE LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.3 Leakage from each Reactor Coolant System Pressure Isolation Valve, specified in Table 3.4-1, shall be equivalent to s 0.5 gpm per 3

nominal inch of valve size up to a maximum of 5 gpm at a Reactor Coolant System pressure = 2215 psig and s 2255 psig.

APPLICABILITY:-MODES 1, 2,

and 3, MODE 4, except valves in the residual heat removal system flow path when in, or during the transition to or from, the residual heat removal mode of operation.

ACTIONS:

a.

With one or more flow paths with leakage from one or more Reactor Coolant System Pressure Isolation Valves greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one closed manual, deactivated automatic, or check valve

and restore the inoperable Reactor Coolant System Pressure Isolation Valve to OPERABLE status within the following 68 hours7.87037e-4 days 0.0189 hours 1.124339e-4 weeks 2.5874e-5 months . Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , b.

Separate entry into the above ACTION is allowed for each flow path, c.

Entry into the applicable ACTIONS for systems made inoperable by an inoperable Reactor Coolant System Pressure Isolation Valve is required.

SURVEILLANCE REQUIREMENTS l

4.4.6.3 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE pursuant to Specification 4.0.5, except that in lieu of any leakage testing requirements required by Specification 4.0.5, each valve shall be demonstrated OPERABLE by verifying leakage to be within its limit':

a.

At least once per 18 months b.

Prior to entering MODE 2 whenever the plant has been in COLD SEUTDOWN for 7 days or more and if leakage testing has not been performed in the previous 9 months.

c.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual

-action or flow through the valve.

Not required to be performed in MODES 3 and 4.

Each valve used to satisfy ACTION a must have been verified to meet the Surveillance Requirement 4.4.6.3 and be in the reactor coolant pressure boundary.

Not required to be performed on' Reactor Coolant System Pressure Isolation Valves located in the Residual Heat Removal flow path when in the shutdown cooling mode of' operation.

SEQUOYAH - UNIT 1 3/4 4-15 Amendment No.

1

EMERGENCY CORE COOLING SYSTEMS (ECCS) 3 /4. 5. 6 SEAL INJECTION FLOW LIMITING CONDITION FOR OPERATION 3.5.6 Reactor coolant pump seal injection flow shall be within limits.

APPLICABILITY: MODES 1, 2,

and 3.

ACTION:

With reactor coolant pump seal injection flow not within limit, adjust manual seal injection throttle valves to give a flow within limit in accordance with surveillance Requirement 4.5.6 within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.6 At least once per 31 days

verify manual seal injection throttle valves are adjusted to give a flow within the emergency core cooling system safety analysis limits.

This surveillance is not required to be performed until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the reactor coolant system pressure stabilizes at a 2215 psig and s 2255 psig.

SEQUOYAH - UNIT 1 3/4 5-12 Amendment No.

REACTOR COOLANT SYSTEM BASES 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION INSTRUMENTATION BACKGROUND GDC 30 of Appendix A to 10 CFR 50 (Ref. 1) requires means for detecting and, to the extent practical, identifying the

. location of the source of reactor coolant system (RCS) i leakage.

Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.

Leakage detection systems must'have the capability to detect significant reactor coolant pressure boundary (RCPB) degradation as soon after occurrence as practical to minimize the potential for propagation to a gross failure.

Thus, an early indication or warning signal is necessary to permit proper evaluation of all UNIDENTIFIED LEAKAGE.

Industry practice has shown that water flow changes of 0.5 to 1.0 gpm can be readily detected in contained volumes by monitoring changes in water level, in flow rate, or in the operating frequency of a pump.

The containment pocket sump used to collect UNIDENTIFIED LEAKAGE is instrumented to alarm for increases of 1.0 gpm in the normal flow rates within one hour.

This sensitivity is acceptable for detecting increases in UNIDENTIFIED LEAKAGE.

The' reactor coolant contains radioactivity that, when released to the containment, can be detected by radiation monitoring instrumentation.

Reactor coolant radioactivity levels will be low during initial reactor startup and for a few weeks thereafter, until activated corrosion products have been formed and fission products appear from fuel element cladding contamination or cladding defects.

4 Instrument sensitivities of 10 pC1/cc radioactivity for 4

particulate monitoring and of 10 pCi/cc radioactivity for gaseous monitoring are practical for these leakage detection systems.

Radioactivity detection systems are included for monitoring both particulate and gaseous activities because of their sensitivities and rapid responses to RCS leakage.

An increase in humidity of the containment atmosphere would indicate release of water vapor to the containment.

Dew point temperature measurements can thus be used to monitor humidity levels of the containment atmosphere as an indicator of potential RCS leakage.

Since the humidity level is influenced by several factors, a quantitative evaluation of an indicated leakage rate by this means may be questionable and should be compared to observed increases in liquid flow into or from the containment sump.

Humidity level monitoring is considered most useful as an indirect alarm or indication to alert the operator to a potential problem.

Hwmidity monitors are not required by this LCO.

SEQUOYAH - UNIT 1 B 3/4 4-4a Amendment No.

REACTOR COOLANT SYSTEM BASES Air temperature and pressure monitoring methods may also be used to infer UNIDENTIFIED LEAKAGE to the containment.

Containment temperature and pressure fluctuate slightly during plant operation, but a rise above the normally indicated range of values may indicate RCS leakage into the containment. The relevance of temperature and pressure measurements are affected by containment free volume and, for temperature, detector location. Alarm signals from these instruments can be valuable in recognizing rapid and sizable leakage to the containment.

Temperature and pressure monitors are not required by this LCO.

APPLICABLE The need to evaluate the severity of an alarm or an SAFETY ANALYSES indication is important to the operators, and the ability to compare and verify with indications from other systems is necessary. The system response times and sensitivities are described in the FSAR (Ref. 3).

Multiple instrument locations are utilized, if needed, to ensure that the transport delay time of the leakage from its source to an instrument location yields an acceptable overall response time.

The safety significance of RCS leakage varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring RCS leakage into the containment area is necessary. Quickly separating the IDENTIFIED LEAKAGE from the UNIDENTIFIED LEAKAGE provides quantitative information to the operators, allowing them to take corrective action should a leakage occur detrimental to the safety of the unit and the public.

Exclusions to the requirements of General Design Criteria 4, for the dynamic effects of the RCS piping, have been utilized based on the leak detection capability to identify leaks before a pipe break would occur.

i' RCS leakage detection instrumentation satisfies Criterion 1 of the NRC Policy Statement.

LCO One method of protecting against large RCS leakage derives from the ability of instruments to rapidly detect extremely small leaks. This LCO requires instruments of diverse monitoring principles to be OPERABLE to provide a high degree of confidence that extremely small leaks are detected in time to allow actions to place the plant in a safe condition, when RCS leakage indicates possible RCPB degradation.

The LCO is satisfied when monitors of diverse measurement j

means are available.

Thus, one containment pocket sump l

monitor, in combination with a gaseous or particulate radioactivity monitor, provides an acceptable minimum.

j SEQUOYAH - UNIT 1 B 3/4 4-4b Amendment No.

1

. REACTOR COOLANT SYSTEM BASES

........-........ =-...........--..............--- -----...................

APPLICABILITY Because of elevated RCS temperature and pressure in MODES 1, 2,

3, and 4, RCS leakage detection instrumentation is required to be OPERABLE.

In MODE 5 or 6, the temperature is to be 5 200'F and pressure is maintained low or at atmospheric pressure.

Since the temperatures and pressures are far lower than those for MODES 1, 2,

3, and 4, the likelihood of leakage and crack propagation are much smaller. Therefore, the requirements of this LCO are not applicable in MODES 5 and 6.

I I

ACTIONS Action a:

With both containment pocket sump monitors inoperable, no other form of sampling can provide the equivalent information; however, the containment atmosphere radioactivity monitor will provide indications of changes in j

leakage. Together with the atmosphere monitor, the periodic 1

surveillance for RCS water inventory balance, Surveillance 4.4.6.2.1, must be performed at an increased frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to provide information that is adequate to detect leakage. A footnote is added allowing that SR 4.4.6.2.1 is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady state operation (stable pressure, temperature, power level, pressurizer and makeup tank levels, makeup, letdown, and RCP seal injection and return flows).

The 12-hour allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

Restoration of the required pocket sump monitor to OPERABLE status within a Completion Time of 30 days is required to regain the function after the monitor's failure. This time is acceptable, considering the frequency and adequacy of the RCS water inventory balance required by Action a.

Action a is modified by a note that indicates that the provisions of LCO 3.0.4 are not applicable. As a result, a MODI change is allowed when the containment sump monitor is inoperable. This allowance is provided because other instrumentation is available to monitor RCS leakage.

If the requirements of Action a cannot be met, the plant must be brought to a MODE in which the requirement 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 the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-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.

SEQUOYAH - UNIT 1 B 3/4 4-4c Amendment No.

REACTOR COOLANT SYSTEM BASES Action b:

With both gaseous and particulate containment atmosphere radioactivity monitoring instrumentation channels inoperable, alternative action is required.

Either grab samples of the containment atmosphere must be taken and analyzed or water inventory balances, in accordance with Surveillance 4.4.6.2.1, must be performed to provide alternate periodic information.

With a sample obtained and analyzed or water inventory balance performed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the reactor may be operated for up to 30 days to allow restoration of at least one containment atmosphere radioactivity monitor.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval provides periodic information that is adequate to detect leakage. A footnote is added allowing that SR 4.4.6.2.1 is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady state operation (stable pressure, temperature, power level, pressurizer and makeup tank levels, makeup, letdown, and RCP seal injection and return flows).

The 12-hour allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

The 30 day Completion Time recognizes at least one other form of leakage detection is available, i

Action b is modified by a note that indicates that the 3

provisions of LCO 3.0.4 are not applicable.

As'a result, a MODE change is allowed when the gaseous and particulate containment atmosphere radioactivity monitor channels are inoperable.

This allowance is provided because other instrumentation is available to monitor for RCS leakage.

If the requirements of Action b cannot be met, th; plant must be brought to a MODE in which the requirement 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 the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-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.

Action c:

j With all required monitors inoperable, no automatic means of monitoring leakage are available, and immediate plant shutdown to a MODE in which the requirement does not apply is required. 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 the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

1 SEQUOYAH - UNIT 1 B 3/4 4-4d Amendment No.

1 1

\\

l l

l

{

l REACTOR COOLANT SYSTEM l

l BASES l

SURVEILLANCE Surveillance 4.4.6.1.a REQUIREMENTS This surveillance requires the performance of a CHANNEL CHECK of the required containment atmosphere radioactivity monitors. The check gives reasonable confidence that the monitors are operating properly.

The frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

]

is based on instrument reliability and is reasonable for detecting off normal conditions.

This surveillance requires the performance of a CHANNEL CALIBRATION for the required containment atmosphere radioactivity monitors.

The calibration verifies the j

accuracy of the instrument string, including the instruments located inside containment.

The frequency of 18 months is a typical refueling cycle and considers channel reliability.

Operating experience has proven that this frequency is acceptable.

This surveillance requires the performance of a CHANNEL FUNCTIONAL TEST on the required containment atmosphere radioactivity monitors.

The test ensures that the monitors can perform their functions in the desired manner.

The test verifies the alarm setpoint and relative accuracy of the instrument string.

The frequency of 92 days considers instrument reliability, and operating experience has shown that it is proper for detecting degradation.

The surveillance frequencies for these tests are specified in Table 4.3-3.

Surveillance 4.4.6.1.b This surveillance requires the performance of a CHANNEL CALIBRATION for the required containment pocket sump monitors. The calibration verifies the accuracy of the instrument string, including the instruments located inside containment.

The frequency of 18 months is a typical refueling cycle and considers channel reliability. Again, operating experience has proven that this frequency is acceptable.

REFERENCES 1.

10 CFR 50, Appendix A, Section IV, GDC 30.

2.

3.

ESAR, Sections 5.2.7 "RCBP Leakage Detection Systems" and 12.2.4 " Airborne Radioactivity Monitoring."

SEQUOYAH - UNIT 1 B 3/4 4-4e Amendment No.

REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE BACKGROUND Components that contain or transport the coolant to or from the reactor core make up the reactor coolant system (RCS).

Component joints'are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from the RCS.

i During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant leakage, through either normal operational wear or mechanical deterioration.

The purpose of the RCS Operational leakage LCO is to limit system operation in the presence of leakage from these sources to amounts that do not compromise safety. This LCO specifies the types and amounts of leakage.

10 CFR 50, Appendix A, GDC 30 (Ref. 1), requires means for detecting and, to the extent practical, identifying the source of reactor coolant leakage.

Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.

The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring reactor coolant leakage into the containment area is necessary. Quickly separating the identified LEAKAGE from the unidentified leakage is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public.

A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100% leaktight.

Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS leakage detection.

This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate' cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded.

The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA).

APPLICABLE Except for primary-to-secondary leakage, the safety analyses l

SAFETY ANALYSES do not address operational leakage.

However, other operational leakage is related to the safety analyses for LOCA; the amount of leakage can affect the probability of such-an event.

The safety analysis for an event resulting in steam discharge to the atmosphere assumes a 1 gpm primary to secondary leakage as the initial condition.

SEQUOYAH - UNIT 1 B 3/4 4-4f Amendment No. 36, 189, 214, 222, 237,

REACTOR COOLANT SYSTEM BASES Primary to secondary leakage is a factor in the dose releases outside containment resulting from a steam line break (SLB) accident.

To a lesser extent, other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR).

The leakage contaminates the secondary fluid.

The FSAR (Ref. 3) analysis for SGTR assumes the contaminated secondary fluid is released via safety valves for up to 30 minutes. Operator action is taken to isolate the affected steam generator within this time period.

The 1 gpm primary to secondary leakage is relatively inconsequential.

The SLB is more limiting for site radiation relcaces.

The safety analysis for the SLB accident assumes 1 gpm primary to secondary leakage in one generator as an initial condition.

The dose consequences resulting from the SLB accident are well within the limits defined in 10 CFR 100 or the staff approved licensing basis (i.e.,

a small fraction of these limits).

Based on the NDE uncertainties, bobbin coil voltage distribution and crack growth rate from the previous inspection, the expected leak rate following a steam line rupture is limited to below 8.21 gpm at atmospheric conditions and 70*F in the faulted loop, which R241 will limit the calculated offsite doses to within 10 percent of the 10 CFR 100 guidelines.

If the projected and cycle distribution of crack indications results in primary-to-secondary leakage greater than 8.21 gpm in the faulted loop during a postulated steam line break event, additional tubes must be removed from service in order to reduce the postulated primary-to-secondary steam line break leakage to below 8.21 gpm.

The RCS operational leakage satisfies Criterion 2 of the NRC Policy Statement.

-LCO RCS operational leakage shall be limited to:

a.

PRESSURE DOgNDARY LEAKAGE No PRESSURL BOUNDARY LEAKAGE is allowed, being indicative of material deterioration. Leakage of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher leakage.

Violation of this LCO could result in continued degradation of the RCPB.

Leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

b.

UNIDENTIFIED LEAKAGE One gpm of UNIDENTIFIED LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment pocket SEQUOYAH - UNIT 1 B 3/4 4-4g Amendment No. 36, 189, 214, 222, 237,

REACTOR COOLANT SYSTEM BASES sump level monitoring equipment can detect within a reasonable time period.

Violation of this LCO could result in continued degradation of the RCPB, if the leakage is from the pressure boundary.

c.

Primary to Secondary Leakage through Any One Steam Generator (SG)

The 150 gallons per day limit on one SG is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line rupture.

If leaked t:. rough many cracks, the cracks are very small, and the above assumption is conservative.

The 150 gallons per day limit incorporated into Surveillance 4.4.6.2.1 is more restrictive than the standard operating leakage limit and is intended to provide an additional margin to accommodate a crack which might grow at a greater than expected rate or R241 unexpectedly extend outside the thickness of the tube support plate.

Hence, the reduced leakage limit, when combined with an effective leak rate monitoring program, provides additional assurance that, should a significant leak be experienced, it will be detected, and the plant shut down in a timely manner, d.

IDENTIFIED LEAKAGE Up to 10 gpm of IDENTIFIED LEAKAGE is considered allowable because leakage is from known sources that do not interfere with detection of UNIDENTIFIED LEAKAGE and is well within the capability of the RCS Makeup System.

IDENTIFIED LEAKAGE includes leakage to the containment from specifically known and located sources, but does not include PRESSURE BOUNDARY LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered leakage).

Violation of this LCO could result in continued degradation of a component or system.

APPLICABILITY In MODES 1, 2,

3, and 4, the potential for reactor coolant PRESSURE BOUNDARY LEAKAGE is greatest when the RCS is pressurized.

In MODES 5 and 6, leakage limits are not required because the reactor coolant pressure is far lower, resulting in l

lower stresses and reduced potentials for leakage.

l SEQUOYAH - UNIT 1 B 3/4 4-4h Amendment No. 36, 189, 214, 222, 237,

1 l

REACTOR COOLANT SYSTEM l

BASES I

l 1

LCO 3/4.4.6.3, "RCS Pressure Isolation Valve (PIV) Leakage,"

measures leakage through each individual PIV and can impact f

this LCO.

Of the two PIVs in series in each isolated line, a

leakage measured through one PIV does not result in RCS leakage when the other is leak tight.

If both valves leak and result in a loss of mass from the RCS, the loss must be l

included in the allowable IDENTIFIED LEAKAGE.

i ACTIONS Action a:

If any PRESSURE BOUNDARY LEAKAGE exists, the reactor must be i

brought to lower pressure conditions to reduce the severity of the leakage and its potential consequences.

It should be noted that leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

The reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

This action reduces the leakage and also reduces the factors that tend to degrade the pressure boundary.

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.

In MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.

Action b; UNIDENTIFIED LEAKAGE, IDENTIFIED LEAKAGE, or primary-to-secondary leakage in excess of the LCO limits must be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

This completion time allows time to verify leakage rates and either identify UNIDENTIFIED LEAKAGE or reduce leakage to within limits l

'before the reactor must be shut down.

This action is i

necessary to prevent further deterioration of the RCPB.

If l

UNIDENTIFIED LEAKAGE, IDENTIFIED LEAKAGE, or primary to secondary leakage cannot be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the reactor must be brought to lower pressure conditions to reduce the severity of the leakage and its potential consequences.

The reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

This action reduces the leakage and also reduces the factors that tend to degrade the pressure boundary.

The allowed completion times are reasonable, based on I

operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

In MODE 5, the pressure stresses i

acting on the RCPB are much lower, and further deterioration I

is much less likely.

SEQUOYAH - UNIT 1 B 3/4 4-41 Amendment No. 36, 189, 214, 222, 237,

REACTOR COOLANT SYSTEM BASES SURVEILLANCE Surveillance-4.4.6.2.1 REQUIREMENTS Verifying RCS leakage to be within the LCO limits ensures the integrity of the RCPB is maintained.

PRESSURE BOUNDARY LEAKAGE would at first appear as UNIDENTIFIED LEAKAGE and can only be positively identified by inspection.

It should be noted that leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

UNIDENTIFIED LEAKAGE and IDENTIFIED LEAKAGE are determined by performance of an RCS water inventory balance.

Primary-to-secondary leakage is also measured by performance of an RCS water inventory balance in conjunction with effluent monitoring within the secondary steam and feedwater systems.

The RCS water inventory balance must be met with the reactor at steady state operating conditions (stable pressure, temperature, power level, pressurizer and makeup tank levels, makeup, letdown, and RCP seal injection and return flows).

Therefore, a footnote is added allowing that this SR is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady state operation. The 12-hour allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

Performance of this surveillance within the 12-hour allowance is required to maintain compliance with the provisions of Specification 4.0.3.

Steady state operation is required to perform a proper inventory balance since calculations during maneuvering are l

not useful.

For RCS operational leakage determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and l

makeup tank levels, makeup and letdown, and RCP seal l'

injection and return flows.

l An early warning of PRESSURE BOUNDARY LEAKAGE or UNIDENTIFIED LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment pocket sump level.

It should be noted that leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

These leakage detection systems are specified in LCO 3/4.4.6.1, " Leakage Detection Instrumentation."

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months frequency is a reasonable interval to trend leakage and recognizes the importance of early leakage detection in the prevention of accidents.

Surveillance 4.4.6.2.2 This surveillance provides the means necessary to determine SG OPERABILITY in an operational MODE.

The requirement to demonstrate SG tube integrity in accordance with the Steam Generator Tube Surveillance Program emphasizes the importance of SG tube integrity, even though this I

surveillance cannot be performed at normal operating conditions.

SEQUOYAH - UNIT 1 B 3/4 4-4j Amendment No. 36, 189, 214, 222, 237,

REACTOR COOLANT SYSTEM 1

BASES:

l 1

i

' REFERENCES.

1.

2.

Regulatory Guide 1.45, May 1973.

q 3.

FSAR, Section 15.4.3.

I

)

i

(

)

1 4

l 1

4

!i l

i SEQUOYAf! - UNIT 1 B 3/4 4-4k Amendment No. 36, 189, 214, 222, 237,

REACTOR' COOLANT SYSTEM BASES

'3/4.4.6.3 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVE LEAKAGE BACKGROUND 10 CFR 50.2, 10 CFR 50.55a (c), and GDC 55 of 10 CFR 50, Appendix A (Refs.

1, 2, and 3), define reactor coolant system (RCS) pressure isolation' valves (PIVs) as any two normally closed valves in series

'within the reactor coolant pressure boundary (RCPB),

which separate the high pressure RCS from an attached low pressure system.

During their lives, these valves can produce varying amounts of reactor coolant leakage through either normal operational wear or mechanical deterioration. The RCS PIV leakage LCO allows RCS j

high pressure operation.when leakage through these j

valves exists in amounts that do not compromise safety.

Although this specification provides a limit on allowable PIV leakage rate, its main purpose is to prevent overpressure failure of the low pressure portions of connecting systems.

The' leakage limit is an indication that the PIVs between the RCS and the co'.ecting systems are degraded or degrading.

PIV leakage could. lead to overpressure of the low pressure piping or components'.

Failure consequences could be a loss of coolant accident (LOCA) outside of containment or an unanalyzed accident that could degrade the ability for low pressure injection.

The basis for this LCO is the 1975 NRC, " Reactor Safety Study," (Ref. 4) that identified potential intersystem LOCAs as a significant contributor to the risk of core melt.

A subsequent study (Ref. 5) evaluated various PIV configurations to determine the l

probability of intersystem LOCAs.

PIVs are provided to isolate the RCS from the following typically connected systems:

a.

Residual. Heat Removal (RHR) System; b.

Safety Injection System; and c.

Chemical and Volume Control System.

The PIVs are listed in Table 3.4-1.

Violation of this LCO could result in continued degradation of a PIV, which could lead to 1

overpressurization of a low pressure system and the loss of the integrity of a fission product barrier.

SEQUOYAH - UNIT 1 B 3/4 4-41 Amendment No.

REACTOR COOLANT SYSTEM BASES APPLICABLE Reference 4 identified potential intersystem LOCAs as SAFETY ANALYSES a significant contributor to the risk of core melt.

The dominant accident sequence in the intersystem LOCA category is the failure of the low pressure portion of the RHR System outside of containment.

The accident is the result of a postulated failure of the PIVs, which are part of the RCPB, and the subsequent pressurization of the RHR System downstream of the PIVs from the RCS.

Because the low pressure portion of the RHR System is typically designed for 600 psig, overpressurization failure of the RHR low pressure line would result in a LOCA outside containment and l

subsequent risk of core melt.

J l

Reference 5 evaluated various PIV configurations, i

leakage testing of the valves, and ope.ational changes

)

to determine the effect on the probability of j

intersystem LOCAs. This study concluded that periodic I

leakage testing of the PIVs can rubstantially reduce the probability of an intersystem LOCA.

RCS PIV leakage satisfies Criterion 2 of the NRC

-Policy Statement.

LCO RCS PIV leakage is IDENTIFIED LEAKAGE into closed i

I systems connected to the RCS.

Isolation valve leakage is usually on the order of drops per minute.

Leakage that increases significantly suggests that something is operationally wrong and corrective action must be taken.

The LCO PIV leakage limit is 0.5 gpm per nominal inch of valve size with a maximum limit of 5 gpm.

The previous criterion of 1 gpm for all valve sizes imposed an unjustified penalty on the larger valves without providing information on potential valve degradation and resulted in higher personnel radiation exposures. A study concluded a leakage rate limit based on valve size was superior to a single allowable value.

Reference 6 permits leakage testing at a lower pressure differential than between the specified maximum RCS pressure and the normal pressure of the I

connected system during RCS operation (the maximum pressure differential) in those types of valves in i

which the higher service pressure will tend to diminish the overall leakage ct$nnal opening.

In such l

cases, the observed rate may be adjusted to the maximum pressure differential by assuming leakage is directly proportional to the pressure differential to the one half power.

SEQUOYAH - UNIT 1 B 3/4 4-4m Amendment No.

n

REACTOR COOLANT' SYSTEM

-BASES APPLICABILITY In MODES 1, 2, 3, and 4, this LCO applies because the PIV leakage potential is greatest when the RCS is pressurized.

In MODE 4, valves in the RHR flow path are not required to meet the requirements of this LCO when in, or during the transition to or from, the RHR mode of operation.

In MODES 5 and 6, leakage limits are not provided because the lower reactor coolant pressure results in a reduced potential for leakage and for a LOCA outside the containment.

ACTIONS Action a:

l The flow path must be isolated.

Action a is modified by a note that the valves used for isolation must meet i

the same leakage requirements as the PIVs and must be l

within the RCPB.

Action a requires that the isolation with one valve must be performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Four hours provides time to reduce leakage in excess of the allowable limit and to icolate the affected system if leakage i

cannot be reduced.

The 4-hour completion time allows the actions and restricts the operation with leaking isolation valves.

J l

The 72-hour completion time after exceeding the limit allows for the restoration of the leaking PIV to 1

OPERABLE status.

This timeframe considers the time required to complete this action and the low probability of a second valve failing during this period.

If leakage cannot be reduced or the system isolated, the plant must be brought to a MODE in which the requirement does not apply.

To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . This Action may reduce the leakage and also reduces the potential for a LOCA outside the containment.

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.

Action b:

l l

Action b provides clarification that each flow path allows separate entry into Action a This is allowed based upon the functional independence of the flow path.

SEQUOYAH - UNIT 1 B 3/4 4-4n Amendment No.

REACTOR COOLANT SYSTEM j

BASES l

Action c:

Action c requires an evaluation of affected systems if a PIV is inoperable.

The leakage may have affected system operability or isolation of a leaking flow path with an alternate valve may have degraded the ability of the interconnected system to perform its safety function.

SURVEILLANCE Surveillance 4.4.6.3 I

REQUIREMENTS Performance of leakage testing on each RCS PIV or isolation valve used to satisfy Action a is required to verify that leakage is below the specified limit and.to identify each leaking valve. The leakage limit of 0.5 gpm per inch of nominal valve diameter up to 5 gpm maximum applies to each valve.

Leakage testing requires a stable pressure condition.

For the two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves.

If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the other valve in series meets the leakage requirement.

In this situation, the protection provided by redundant valves would be lost.

Testing is to be performed every 18 months, a typical refueling cycle, if the plant does not go into MODE 5 for at least 7 days.

The 18 month frequency is consistent with 10 CFR 50.55a(g) (Ref. 7) as contained in the Inservice Testing Program, is within frequency allowed by the American Society of Mechanical Engineers (ASME) Code,Section XI (Ref. 6), and is based on the need to perform such surveillances under the conditions that apply during an outage and the potential for an unplanned transient if the surveillances were performed with the reactor at power.

In addition, testing must be performed once after the valve has been opened by flow or exercised to ensure tight reseating.

PIVs disturbed in the performance of this surveillance should also be tested unless documentation shows that an infinite testing loop cannot practically be avoided. Testing must be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the valve has been reseated. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a reasonable and

-practical time limit for performing this test after opening:or reseating a valve.

SEQUOYAH - UNIT 1 B 3/4 4-4o Amendment No.

REACTOR COOLANT SYSTEM y

BASES l

The leakage limit is to be met at the RCS pressure associated with MODES 1 and 2.

This permits leakage testing at high differential pressures with stable conditions not possible in the MODES with lower pressures.

Entry into MODES 3 and 4 is allowed to establish the necessary differential pressures and stable conditions to allow for performance of this surveillance. The note that allows this provision is complementary to the frequency of prior to entry into MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months.

In addition, this surveillance is not required to be performed on the RHR System when the RHR System is aligned to the RCS in the shutdown cooling mode of operation.

PIVs contained in the RHR shutdown cooling flow path must be leakage rate tested after RHR is secured and stable unit conditions and the necessary differential pressures are established.

l REFERENCES 1.

2.

10 CFR 50.55a(c).

3.

10 CFR 50, Appendix A, Section V, GDC 55.

4.

WASH-1400 (NUREG-75/014), Appendix V, October 1975.

5.

NUREG-0677, May 1980.

6.

ASME, Boiler and Pressure Vessel Code,Section XI.

7.

SEQUOYAH - UNIT 1 B 3/4 4-4p Amendment No.

i REACTOR COOLANT SYSTEM

- BASES 3/4.4,7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion s

of the Reactor Coolant System is' minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent.

Corrosion studies show that operation may be continued with contaminant concentration i

levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time; intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

1 October 11, 1995 SEQUOYAH - UNIT 1 B 3/4 4-4q Amendment No. 36, 189, 214 l

1 i

EMERGENCY CORE COOLING SYSTEM BASES 3/4.5.6 SEAL INJECTION FLOW BACKGROUND The function of the seal injection throttle valves during an accident is similar to the function of the ECCS throttle valves in that each restricts flow from the centrifugal charging pump header to the Reactor Coolant System (RCS).

The restriction on reactor coolant pump (RCP) seal j

injection flow limits the amount of ECCS flow that I

would be diverted from the injection path following an 1

accident.

This limit is based on safety analysis assumptions that are required because RCP seal injection flow is not isolated during safety j

injection.

1 l

APPLICABLE f.11 ECCS subsystems are taken credit for in the large

]

SAFETY ANALYSES' break loss of coolant accident (LOCA) at full power (Ref. 1).

The LOCA analysis establishes the minimum flow for the ECCS pumps.

The centrifugal charging pumps are also credited in the small break LOCA analysis. This analysis establishes the flow and discharge head at the design point for the centrifugal charging pumps. The steam generator tube rupture and 1

main steam line break event analyses also credit the i

centrifugal charging pumps, but are not limiting in i

their design.

Reference to these analyses is made in assessing changes to the Seal Injection System for evaluation of their effects in relation to the acceptance limits in these analyses.

This LCO ensures that seal injection flow will be sufficient for RCP seal integrity but limited so that the ECCS trains will be capable of delivering sufficient water to match boiloff rates soon enough to minimize uncovering of the core follcwing a large LOCA.

It also ensures that the centrifugal charging pumps will deliver sufficient water for a small LOCA and sufficient boron to maintain the core suberitical.

For smaller LOCAs, the charging pumps alone deliver sufficient fluid to overcome the loss and maintain RCS inventory.

Seal injection flow satisfies Criterion 2 of the NRC Policy Statement.

SEQUOYAH - UNIT 1 B 3/4 5-4 Amendment No.

i

r EMERGENCY CORE COOLING SYSTEM BASES LCO The !.ntent of the LCO limit on seal injection flow is to make sure that flow through the RCP seal water injection line is low enough to ensure that sufficient centrifugal charging pump injection flow is directed to the RCS via the injection points (Ref. 2).

The LCO is not strictly a flow limit, but rather a f1 w limit based on a flow line resistance.

In order to esttblish the proper flow line resistance, a pressure and flow must be known.

The flow line resistance is established by adjusting the RCP seal injection needle valves to provide-a total seal injection flow in the acceptable region of Figure B 3.5.6-1.

The centrifugal charging pump discharge header pressure remains essentially constant I

through all the applicable MODES of this LCO.

A reduction in RCS pressure would result in more flow being diverted to the RCP seal injection line than at normal operating pressure.

The valve settings established at the prescribed centrifugal charging pump discharge header pressure result in a conservative valve position'should RCS pressure decrease.

The flow limits established by Figure B 3.5.6-1~are consistent with the accident l

analysis.

The limits on seal injection flow must be met to render the SCCS OPERABLE.

If these conditions are not i

met, the ECCS flow will not be as assumed in the accident analyses.

I 1

APPLICABILITY In MODES 1, 2, and 3, the seal injection flow limit is dictated by ECCS flow requirements, which are I

specified for MODES 1, 2,

3, and 4.

The seal injection flow limit is not applicable for MODE 4 and j

lower, however, because high seal injection flow is l

less critical as a result of the lower initial RCS pressure and decay heat removal requirements in these MODES.

Therefore, RCP seal injection flow must be limited in MODES 1, 2, and 3 to ensure adequate ECCS performance, i

SEQUOYAH - UNIT 1 B 3/4 5-5 Amendment No.

I

l EMERGENCY CORE COOLING SYSTEM BASES ACTION

-With the seal injection flow exceeding its limit, the amount of charging flow available to the RCS may be reduced.

Under this condition, action must be taken to restore the flow to below its limit.

The operator has 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from the time the flow is known to be above the limit to correctly position the manual valves and thus be in compliance with the accident analysis.

The completion time minimizes the potential

. exposure of the plant to a LOCA with insufficient injection flow and provides a reasonable time to restore seal injection flow within limits.

This time is conservative with respect to the completion times of other ECCS LCOs; it is based on operating i

experience and is sufficient for taking corrective actions by operations personnel.

When the actions cannot be completed within the required completion time, a controlled shutdown must be initiated.

The completion time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for reaching MODE 3 from MODE 1 is a. reasonable time for a i

controlled shutdown, based on operating experience and normal cooldown rates, and does not challenge plant safety systems or operators.

Continuing the plant shutdown from MODE 3, an additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is a reasonable time, based on operating experience and j

normal cooldown rates, to reach MODE 4, where this LCO l

is no longer applicable.

1 SURVEILLANCE Surveillance 4.5.6 REQUIREMENTS Verification every 31 days that the manual seal l

injection throttle valves are adjusted to give a flow l

within the limit ensures that proper manual seal injection throttle valve position, and hence, proper seal injection flow, is maintained.

The differential pressure that is above the reference minimum value is established between the charging header (PT 62-92) and the RCS, and total seal injection flow is verified to be within the limits determined in accordance with the ECCS safety analysis (Ref. 3).

The seal water injection flow limits are shown in Figure B 3.5.6-1.

The frequency of 31 days is based on engineering judgment and is consistent with other ECCS valve surveillance frequencies. The frequency has proven to be acceptable through operating experience.

The requirements for charging flow vary widely

)

according to plant status and configuration. When j

charging flow is adjusted, the positions of the air-operated valves, which control charging flow, are SEQUOYAH - UNIT 1 B 3/4 5-6 Amendment No.

1

EMERGENCY CORE COOLING SYSTEM BASES adjusted to balance the flows through the charging header and through the seal injection header to ensure that the seal injection flow to the RCPs is maintained between 8 and 13 gpm per pump.

The reference minimum differential pressure across the seal injection needle valves ensures that regardless of the varied settings of the charging flow control valves that are required to support optimum charging flow, a reference test condition can be established to ensure that flows across the needle valves are within the safety analysis.

The values in the safety analysis for this reference set of conditions are calculated based on conditions during power operation and they are correlated to the minimum ECCS flow to be maintained under the most limiting accident conditions.

As noted, the surveillance is not required to be performed until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the RCS pressure has stabilized within a t 20 psig range of normal operating pressure.

The RCS pressure requirement is specified since this configuration will produce the required pressure conditions necessary to assure that the manual valves are set correctly.

The exception is limited to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to ensure that the surveillance is timely.

Performance of this surveillance within the 4-hour allowance is required to maintain compliance with the provisions of Specification 4.0.3.

REFERENCES 1.

FSAR, Chapter 6.3 " Emergency Core Cooling System" and Chapter 15.0 " Accident Analysis".

2.

3.

Westinghouse Electric Company Calculation CN-FSE-99-48 SEQUOYAR - UNIT 1 B 3/4 5-7 Amendment No.

o.

N tn e

m d

0 n

5 e

mA 5

4 s

t 0

)

i m

4 m

1 p

- i

6. L g

(

w

.E

5. o s L w

s 3l

<BN o

8 l

F

. A O F

B

. TI 5

n n

.. PG o

/4 Eo

EE i

3 i

t Rt

CR c

B s

c

e. C e

Ue

- A n

j j

Gn i

3 l

I a

i e

Fl 5

a 3 S e

S 0

3

/

I 1

T NU 52 H

0 0

0 A

5 0

5 Y

2 2

1 1

OU EWE 5 ea E* $U HQ. Eabg EE.En.

EO Q

ES

INDEX 1

I DEFINITIONS i

SECT.LQE EAGE 1

1.0-DEFINITIONS l

I 1.1 ACTION 1-1 1.2

. AXIAL FLUX-' DIFFERENCE 1-1 4

' 1.3 BYPASS LEAKAGE PATH.

1-1

.1.4 CHANNEL CALIBRATION 1-1 R63 1.5 CHANNEL CHECK.

1-1

-1.6 CHANNEL FUNCTIONAL TEST 1-2 1.7 CONTAINMENT INTEGRITY 1-2 1.8 CONTROLLED LEAKAGE.(Deleted) 1-2 l

1.9 CORE ALTERATION 1-2 R63 l'.10 CORE OPERATING LIMITS REPORT 1-2 l

1.11 DOSE EQUIVALENT I-131 1-3 1.12 E - AVERAGE DISINTEGRATION ENERGY 1-3 1.13 ENGINEERED SAFETY FEATURE RESPONSE TIME 1-3 R146 1.14.' FREQUENCY NOTATION _

1-3 1.15 GASEOUS RADWASTE TREATMENT SYSTEM 1-3 1.16 IDENTIFIED LEAKAGE 1-4 1.17 MEMBERS OF THE PUBLIC 1-4 i

1.18 OFFSITE DOSE CALCULATION MANUAL 1-4 1.19 OPERABLE OPERABILITY 1-4

)

1-5 1.20 OPERATIONAL MODE - MODE 1.21 PHYSICS TESTS 1-5

'1.22 PRESSURE BOUNDARY LEAKAGE 1-5 1.23-PROCESS CONTROL PROGRAM 1-5 SEQUOYAH - UNIT 2 I

Amendment No, 63, 146,

k INDEX i

LIMITING CONDITIONS FOR OPERATION AND' SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER DPERATION 3/4 4-1 Hot-Standby.

3/4 4-2 Hot Shutdown 3/4 4-3 Cold Shutdown.

3/4 4-5 3 /4.'4. 2 SAFETY VALVES - SHUTDOWN.

3/4 4-6 3/4.4.3 SAFETY AND RELIEF VALVES - OPERATING Safety, Valves Operating.

3/4 4-7 Relief Valves Operating.

3/4 4.................

~ 3/4.4.4 PRESSURIZER' 3/4 4-9 3/4.4.5 STEAM GENERATORS 3/4 4-10 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE

)

Leakage Detection Instrumentation 3/4 4-17 l

operational Leakage 3/4 4-18 Reactor Coolant System Pressure Isolation Valve Leakage 3/4 4-19 l

3/4.4.7 CHEMISTRY 3/4 4-21 J

3/4.4.8 SPECIFIC ACTIVITY 3/4 4-24 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System 3/4 4-28 Pressurizer 3/4 4-31 lR138 3/4.4.'10 DELETED 3/4 4-32 lR198 1

3/4.4.11 REACTOR COOLANT SYSTEM HEAD VENTS 3/4 4-33 lR138 3/4.4.12 OVERPRESSURE PROTECTION SYSTEMS 3/4 4-34 lR147 SEQUOYAH - UNIT 2 VI Amendment No. 106, 120, 138, 147, 198,

p --

.f l

INDEX I

i LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3 /4. 5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

~3/4.5.1 ACCUMULATORS COLD LEG INJECTION ACCUMULATORS 3/4 5-1 DELETED 3/4 5-3 1

3/4.5.2 ECCF SUBSYSTEMS -

T,y, Greater Than or Equal to 350*F 3/4 5-4 3/4.5.3 ECCS SUBSYSTEMS - T,,, LESS THAN. 350*F 3/4 5-8 R131 3/4.5.4 DELETED 3/4 5-10 3/4.5.5 REFUELING WATER STORAGE TANK 3/4 5-11 3/4.5.6 SEAL INJECTION FLOW 3/4 5-12 l

l 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY.

3/4 6-1 1

3/4 6-2 jR167 SECONDARY CONTAINMENT BYPASS LEAKAGE CONTAINMENT AIR LOCKS 3/4 6-7 INTERNAL PRESSURE 3/4 6-9 AIR TEMPERATURE 3/4 6-10 3/4 6-11 CONTAINMENT VESSEL STRUCTURAL INTEGRITY SHIELD BUILDING STRUCTURAL INTEGRITY 3/4 6-12 EMERGENCY GAS TREATMENT SYSTEM (CLEANUP SUBSYSTEM) 3/4 6-13 CONTAINMENT VENTILATION SYSTEM 3/4 6-15 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS

' CONTAINMENT SPRAY SUBSYSTEMS 3/4 6-16 lR140 3/4 6-16b lR;l LOWER CONTAINMENT VENT COOLERS l

l SEQUOYAH - UNIT 2 VII Amendment Nos. 59, 61, 131, 140, l

167,

l INDEX BASES 1

l SECTION PAGE l

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE B 3/4 4-4 3/4.4.7 CHEMISTRY B 3/4 4-5 l

3/4.4.0 SPECIFIC ACTIVITY B S/4 4-5 3/4.4.9 PRESEURE/ TEMPERATURE LIMITS

. B 3/4 4-6 3/4.4.10 STRUCTURAL INTEGRITY B 3/4 4-14 1

l 3/4.4.11 REACTOR COOLANT SYSTEM HEAD VENTS B 3/4 4-15 R147 3/4.4.12 OVERPRESSURE PROTECTION SYSTEMS B 3/4 4-15 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS B 3/4 5-1 3/4.5.4 BORON INJECTION SYSTEM.

B 3/4 5-2 3/4.5.5 REFUELING WATER STORAGE TANK B 3/4 5-2 3/4.5.6 SEAL INJECTION FLOW B 3/4 5-4 l

l l

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS B 3/4 6-3 j

3/4.6.3 CONTAINMENT ISOLATION VALVES B 3/4 6-3 3/4.6.4 COMBUSTIBLE GAS CONTROL B 3/4 6-4 l

3/4.6.5 ICE CONDENSER B 3/4 6-4 3/4.6.6 VACUUM RELIEF VALVES B 3/4 6-6 lR188 3/4.7 PLAtTI' SYSTEMS 3/4.7.1 TURBINE CYCLE B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.

B 3/4 7-3 1

3/4.7.3 COMPONENT COOLING WATER SYSTEM B 3/4 7-3 l

SEQUOYAH - UNIT 2 XIII Amendment No. 147, 188, 1

l

DEFINITIONS CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be:

a.

Analog channels - the injection of a simulated signal into the j

channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions, b.

Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

c.

Digital channels - the injection of a simulated signal into the channel as close to the sensor input to the process racks as R132 practicable to verify OPERABILITY including alarm and/or trip functions.

\\

CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

a.

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

1)

Capable of being closed by an OPERABLE containment automatic isolation valve system, or 2)

Closed by manual valves, blind flanges, or deactivated auto-mati2 ' valves secured in their closed positions, except for R193 valves that are open under administrative control as permitted by Specification 3.6.3.

b.

All equipment hatches are closed and sealed, c.

Each air lock is in compliance with the requirements of lR117 i

Specification 3.6.1.3, d.

The' containment leakage rates are within the limits of Specification 4.6.1.1.c, and R167 e.

The sealing mechanism associated with each penetration

-(e.g.,

welds, bellows, or O-rings) is OPERABLE, and d.

Secondary containment bypass leakage is within the limits of Specification 3.6.1.2.

CONTROLLED LEAKAGE 1.8 This definition has been deleted.

CORE ALTERATION 1,9 CORE ALTERATION shall be the movement of any fuel, sources, reactivity R191 control components, or other components affecting reactivity within the reactor vessel with the head removed and fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT 1.10 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle.

These cycle-specific core operating limits shall be determined for each reload R146 cycle in accordance with specification 6.9.1.14.

Unit operation within these operating limits is addressed in individual specifications.

SEQUOYAH - UNIT 2 1-2 Amendment Nos. 63, 117, 132, 146, 167, 191, 193,

i 1

i DEFINITIONS IDENTIFIED LEAKAGE 1.16 IDENTIFIED LEAKAGE shall be:

lR146 a.

Leakage, such as that from pump seals or valve packing (except i

reactor coolant pump seal injeccion or leakoff) that is captured and conducted to collection systems or a sump or collecting tank, I

or b.

Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or c.

Reactor coolant system leakage through a steam generator to the I

secondary system.

I MEMBERS OF THE PUBLIC 1.17 MEMBERS OF THE PUBLIC means an individual in a controlled or unrestricted area. However, an individual is not a member of the public during any period R165 in which the individual receives an occupational dose.

l OFFSITE DOSE CALCULATION MANUAL 1.18 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology R134 and parameters used in the calculation of offsite doses resulting from radio-active gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduct of the i

Radiological Environmental Monitoring Program.

The ODCM shall also contain (1) l the Radioactive Effluent Controls and Radiological Environmental Monitoring j

Programs required by Section 6.8.4 and (2) descriptions of the information that lR169 should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.6 lR159 and 6.9.1.8.

OPERABLE - OPERABILITY 1.19 A system, subsystem, train, or component or device shall be OPERABLE or l

have OPERABILITY when it is capable of performing its specified function (s),

and when all necessary attendant instrumentation, controls, a normal and an emergency electrical power source, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, com-ponent or device to perform its function (s) are also capable of performing their related support function (s).

l SEQUOYAH - UNIT 2 1-4 Amendment Nos. 63, 134, 146, 159, 165, 169,

l DEFINITIONS l

t SOLIDIFICATION

'1.32~' Deleted.

R146

' SOURCE CHECK 1.33 Deleted.

R146 STAGGERED TEST BASIS 1.34 A STAGGERED TEST BASIS shall consist of:

lR146 a.

A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals, b.

The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.

THERMAL POWER E1.35 THERNAL POWER shall be the total reactor core heat transfer rate to the lR146 reactor coolant.

UNIDENTIFIED LEAKAGE l'36 UNIDENTIFIED LEAKAGE shall be all leakage (except reactor coolant pump seal water injection or leakoff) that is not IDENTIFIED LEAKAGE.

UNRESTRICTED AREA 1.37 An UNRESTRICTED AREA shall be any area, at or beyond the site boundary to lR146 which access is not controlled by.the licensee for purposes of protection of l

individuals from exposure to radiation and radioactive materials or any area R63 within the site boundary used for residential quarters or industrial, commer-l cial, institutional, and/or recreational purposes.

l l

J SEQUOYAH ' UNIT 2 1-7 Amendmemt Nos. 63, 134, 146,

,~

l REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION INSTRUMENTATION l

LIMITING CONDITION FOR OPERATION l

-3.4.6.1 The following Reactor Coolant System leakage detection instrumentation l

shall be OPERABLE:

a.

One lower containment atmosphere radioactivity monitor (gaseous or particulate), and b.

One containment pocket sump level monitor.

APPLICABILITY: MODES 1, 2,

3 and 4.

ACTIONS:

With both containment pocket sump monitors inoperable, operation may a.

continue for up to 30 days provided SR 4.4.6.2.1 is performed once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months *; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The provisions of Specification 3.0.4 are not applicable.

b.

With both lower containment atmosphere radioactivity monitors j

(gaseous and particulate) inoperable, operation may continue for up to 30 days provided grab samples of the lower containment atmosphere are analyzed once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or SR 4.4.6.2.1 is performed once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months *; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

The provisions of Specification 3.0.4 are not applicable.

c.

With both containment pocket sump monitors and both lower containment atmosphere radioactivity monitors inoperable, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.6.1 The leakage detection instrumentation shall be demonstrated OPERABLE l

by:

a.

Performance of the lower containment atmosphere gaseous and particulate monitor CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3, and b.

Performance of containment pocket sump level monitor CHANNEL l

CALIBRATION at least once per 18 months.

Surveillance performance not required until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

SEQUOYAH - UNIT 2 3/4 4-17 Amendment No.

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 13.4.6.2 Reactor Coolant System leakage shall be limited to:

i a.

No PRESSURE BOUNDARY LEAKAGE, b.

1 GPM UNIDENTIFIED LEAKAGE, c.

150 gallons per day of primary-to-secondary leakage through any one lR213 steam generator, and d.

10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System.

APPLICABILITY: MODES 1,.2, 3 and 4 ACTION:

a.

With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , b.

With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage l

rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System-leakages shall be verified to be within each of the above limits by performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .*

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

4.4.6.2.2 Verify steam generator tube integrity is in accordance with the requirements-of Technical Specification 3/4.4.5, " Steam Generators."

i

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

SEQUOYAH - UNIT 2 3/4 4-18 Amendment No. 211, 213, I

y

e-l REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVE LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.3 Leakage from each Reactor Coolant System Pressure Isolation Valve, specified in Table 3.4-1, shall be equivalent to s 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm at a Reactor Coolant System pressure a 2215 psig and s 2255 psig.

APPLICABILITY

MODES 1, 2,

and 3, MODE 4, except valves in the residual heat removal system flow path when in, or during the transition to or from, the residual heat removal mode of operation.

1 ACTIONS:

a.

With one or more flow paths with leakage from one or more Reactor Coolant System Pressure Isolation Valves greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one closed manual, deactivated automatic, or check valve

and restore the inoperable' Reactor Coolant System Pressure Isolation Valve to OPERABLE status.within the following 68 hours7.87037e-4 days 0.0189 hours 1.124339e-4 weeks 2.5874e-5 months . Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD'SHUTDORN within j

the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

i b.

Separate entry into the above ACTION is allowed for each flow path.

c.

Entry into the applicable ACTIONS for systems made inoperable by an inoperable Reactor Coolant System Pressure Isolation Valve is j

required.

]

SURVEILLANCE REQUIREMENTS 4.4.6.3 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE pursuant to Specification 4.0.5, except that in lieu of any leakage testing requirements required by Specification 4.0.5, each valve shall be demonstrated OPERABLE by ve'ifying leakage to be within its limit :

a.

At least once per 18 months b.

Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 7 days or more and if leakage testing has not been performed in the previous 9 months.

c.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or' flow through the valve.

Not required to be performed in MODES 3 and 4.

Each valve used to satisfy ACTION a must have been verified to meet the Surveillance Requirement 4.4.6.3 and be in the reactor coolant pressure boundary.

Not required to be performed on Reactor Coolant System Pressure Isolation Valves. located in the Residual Heat Removal flow path when in the shutdown cooling' mode of operation.

SEQUOYAH - UNIT 2 3/4 4-19 Amendment No.

F.

EMERGENCY CORE COOLING SY3TEMS (ECCS)-

l

' 3 /4 '. 5. 6 SEAL INJECTION FLOW LIMITING CONDITION FOR OPERATION l

3.5.6 Reactor coolant pump seal injection flow shall be within limits.

' APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With reactor' coolant pump seal injection flow not within limit, adjust manual seal ~ injection throttle valves to give a flow within limit in accordance with I

Surveillance Requirement 4.5.6 within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l SURVEILLANCE REQUIREMENTS-I 4.5.6 ~At least once per 31 days

verify manual seal injection. throttle valves are adjusted to give a flow within the emergency core cooling system safety analysis limits.

l i

This surveillance is not required to be performed until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the reactor coolant system pressure stabilizes at = 2215 psig and s 2255 psig.

l SEQUOYAH - UNIT 2 3/4 5-12 Amendment No.

l

REACTOR COOLANT SYSTEM BASES 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1-LEAKAGE DETECTION INSTRUMENTATION BACKGROUND GDC 30 of Appendix A to 10 CFR 50 (Ref. 1) requires means for detecting and, to the extent practical, identifying the location of the source of reactor coolant system (RCS) leakage.

Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.

Leakage detection systems must have the capability to detect significant reactor coolant pressure boundary (RCPB) degradation as soon after occurrence as practical to j

minimize the potential for propagation to a gross failure.

Thus, an early indication or warning signal is necessary to permit proper evaluation of all UNIDENTIFIED LEAKAGE.

Industry practice has shown that water flow changes of 0.5 to 1.0 gpm can be readily detected in contained volumes by monitoring changes in water level, in flow rate, or in the operating frequency of a pump. The containment pocket sump used to collect UNIDENTIFIED LEAKAGE is instrumented to alarm for increases of 1.0 gpm in the normal flow rates within one hour.

This sensitivity is acceptable for detecting increases in UNIDENTIFIED LEAKAGE.

J The reactor coolant contains radioactivity that, when released to the containment, can be detected by radiation j

monitoring instrumentation. Reactor coolant radioactivity levels will be low during initial reactor startup and for a few weeks thereafter, until activated corrosion products have been formed and fission products appear from fuel j

element cladding contamination or cladding defects.

4 Instrument sensitivities of 10 Ci/cc radioactivity for 4

particulate monitoring and of 10 pCi/cc radioactivity for gaseous monitoring are practical for these leakage detection systems.

Radioactivity detection systems are included for monitoring both particulate and gaseous activities because of their sensitivities and rapid responses to RCS leakage.

An increase in humidity of the containment atmosphere would indicate release of water vapor to the containment.

Dew point temperature measurements can thus be used to monitor humidity levels of the containment atmosphere as an indicator of potential RCS leakage.

Since the humidity level is influenced by several factors, a quantitative evaluation of an indicated leakage rate by this means may be questionable and should be compared to observed increases in liquid flow into or from the containment sump.

l Humidity level monitoring is considered most useful as an indirect alarm or indication to alert the operator to a potential problem.

Humidity monitors are not require:, 5y this LCO.

SEQUOYAH - UNIT 2 8 3/4 4-4 Amendment No.

L

J REACTOR COOLANT SYSTEM BASES l

...............--=====...----...... --......--........--....--.....----.....

Air temperature and pressure monitoring methods may also be used to infer UNIDENTIFIED LEAKAGE to the containment.

Containment temperature and pressure fluctuate slightly during plant operation, but a rise above the normally indicated range of values may indicate RCS leakage into the containment. The relevance of temperature and pressure measurements are affected by containment free volume and, for temperature, detector location. Alarm signals from these instruments can be valuable in recognizing rapid and sizable leakage to the containment.

Temperature and i

pressure monitors are not required by this LCO.

i APPLICABLE The need to evaluate the severity of an alarm or an SAFETY ANALYSES indication is important to the operators, and the ability to j

compare and verify with indications from other systems is necessary. The system response times and sensitivities are 1

described in the FSAR (Ref. 3).

Multiple instrument locations are utilized, if needed, to ensure that the transport delay time of the leakage from its source to an instrument location yields an acceptable overall response time.

j i

The safety significance of RCS leakage varies widely I

depending on its source, rate, and duration.

Therefore, j

detecting and monitoring RCS leakage into the containment I

area is necessary.

Quickly separating the IDENTIFIED LEAKAGE from the UNIDENTIFIED LEAKAGE provides quantitative information to the operators, allowing them to take corrective action should a leakage occur detrimental to the safety of the unit and the public.

Exclusions to the requirements of General Design Criteria 4, for the dynamic effects of the RCS piping, have been utilized based on the leak detection capability to identify leaks before a pipe break would occur.

RCS leakage detection instrumentation satisfies Criterion 1 of the NRC Policy Statement.

LCO One method of protecting against large RCS leakage derives from the ability of instruments to rapidly detect extremely small leaks. This LCO requires instruments of diverse monitoring principles to be OPERABLE to provide a high degree of confidence that extremely small leaks are detected in time to allow actions to place the plant in a safe condition, when RCS leakage indicates possible RCPB j

degradation.

Tbs LCO is satisfied when monitors of diverse measurement means are available.

Thus, one con"ainment pocket sump monitor, in combination with a gaseous or particulate radioactivity monitor, provides an acceptaole minimum.

SEQUOYAH - UNIT 2 B 3/4 4-4a Amendment No.

f l-l l

REACTOR COOLANT SYSTEM l

BASES l

l APPLICABILITY Because of elevated RCS temperature and pressure in MODES 1, l

2, 3, and 4, RCS leakage detection instrumentation is required to be OPERABLE.

l In MODE 5 or 6,. the temperature is to be s 200*r and pressure is maintained low or at atmospheric pressure.

Since the temperatures and pressures are far lower than those for MODES 1, 2, 3, and 4, the likelihood of leakage and crack propagation are much smaller.

Therefore, the requirements of this LCO are not applicable in MODES 5 and 6.

ACTIONS Action a:

With both containment pocket sump monitors inoperable, no other form of sampling can provide the equivalent information; however, the containment atmosphere radioactivity monitor will provide indications of changes in leakage. Together with the atmosphere monitor, the periodic surveillance for RCS water inventory balance, Surveillance 4.4.6.2.1, must be performed at an increased frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to provide information that is adequate to detect leakage. A footnote is added allowing that SR 4.4.6.2.1 is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady state operation (stable prersure, temperature, power level, pressurizer and makeup tark levels, makeup, letdown, and RCP seal injection and return flows).

The 12-hour allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

Restoration of the required pocket sump monitor to OPERABLE status within a Completion Time of 30 days is required to regain the function after the monitor's failure.

This time is acceptable, considering the frequency and adequacy of the RCS water inventory balance required by Action a.

Action a is modified by a note that indicates that the provisions of LCO 3.0.4 are not applicable. As a result, a MODE change is allowed when the containment sump monitor is inoperable. This all ance is provided because other instrumentation is available to monitor RCS leakage.

If the requirements of Action a cannot be met, the plant must be brought to a MODE in which the requirement 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 the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-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.

SEQUOYAH - UNIT 2 B 3/4 4-4b Amendment No.

REACTOR COOLANT SYSTEM BASES Action b:

With both gaseous and particulate containment atmosphere radioactivity monitoring instrumentation channels ~

inoperable, alternative action is required.

Either grab samples of the containment atmosphere must be taken and analyzed or water inventory balances, in accordance with Surveillance 4.4.6.2.1, must be performed to provide alternate periodic information.

With a sample obtained and analyzed or water inventory balance performed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the reactor may be operated for up to 30 days to allow restoration of at least one containment atmosphere radioactivity monitor.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval provides periodic information that is adequate to detect leakage. A footnote is added allowing that SR 4.4.6.2.1 is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady state operation (stable pressure, temperature, power level, pressurizer and makeup tank levels, makeup, letdown, and RCP seal injection and return flows).

The 12-hour allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

The 30 day Completion Time recognizes at least one other form of leakage detection is available.

Action b is modified by a note that indicates that the provisions of LCO 3.0.4 are not applicable. As a result, a MODE change is allowed when the gaseous and particulate conta3 ment atmosphere radioactivity monitor channels are inope'able.

This allowance is provided because other instrumentation is available to monitor for RCS leakage.

1 If the requirements of Action b cannot be met, the plant must be brought to a MODE in which the requirement does net 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 the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-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.

)

Action c:

With all required monitors inoperable, no automatic means of monitoring leakage are available, and immediate plant shutdown to a MODE in which the requirement does not apply is required.

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 the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SEQUOYAH - UNIT 2 B 3/4 4-4c Amendment No.

REACTOR COOLANT SYSTEM BASES

..w.........................................................................

SURVEILLANCE Surveillance 4.4.6.1.a REQUIREMENTS This surveillance requires the performance of a CHANNEL CHECK of the required containment atmosphere radioactivity monitors.

The check gives reasonable' confidence that the monitors are operating properly.

The frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is based on instrument reliability and is reasonable for detecting off normal conditions.

l This surveillance requires the performance of a CHANNEL l

CALIBRATION for the required containment atmosphere radioactivity monitors. The calibration verifies the J

accuracy of the instrument string, including the instruments

)

located inside containment. The frequency of 18 months is a l

typical refueling cycle and considers channel reliability.

Operating experience has proven that this frequency is j

acceptable.

This surveillance requires ths performance of a CHANNEL FUNCTIONAL TEST on the required containment atmosphere radioactivity monitors.

The test ensures that the monitors can perform their functions in the desired manner.

The test verifies the alarm setpoint and relative accuracy of the instrument string.

The frequency of 92 days considers instrument reliability, and operating experience has shown that it is proper for detecting degradation.

The surveillance frequencies for these tests are specified in Table 4.3-3.

Surveillance 4.4.6.1.b This surveillance requires the performance of a CHANNEL CALIBRATION for the required containment pocket sump monitors. The calibration verifies the accuracy of the i

instrument string, including the instruments located inside containment.

The frequency of 18 months is a typical refueling cycle and considers channel reliability. Again, operating experience has proven that this frequency is acceptable.

REFERENCES 1.

10 CFR 50, Appendix A, Section IV, GDC 30.

2.

3.

FSAR, Sections 5.2.7 "RCBP Leakage Detection Systems" and 12.2.4 " Airborne Radioactivity Monitoring."

SEQUOYAH - UNIT 2 B 3/4 4-4d Amendment No.

i REACTOR COOLANT SYSTEM BASES I

i 3/4'4.6.2 OPERATIONAL LEAKAGE j

BACKGROUND Components that contain or transport the coolant to or from the reactor core make up the reactor coolant system (RCS).

Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from j

the RCS.

i j

During plant life, the joint and valve interfaces can I

produce varying amounts of reactor coolant leakage, through either normal' operational wear or mechanical deterioration.

~The purpose of the RCS Operational leakage LCO is to limit system operation in the presence of leakage from these sources to amounts that do not compromise safety. This LCO specifies the types'and amounts of leakage.

10 CFR 50, Appendix A, GDC 30 (Ref. 1), requires means for detecting and, to the extent practical, identifying the source of reactor coolant leakage.

Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.

The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration.

Therefore, detecting and monitoring reactor coolant leakage into the containment area is necessary. Quickly separating the identified LEAKAGE from the unidentified leakage is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public.

A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100% leaktight.

Leakage from these systems should be detected, located, and isolated from the~ containment atmosphere, if possible, to not interfere with RCS leakage detection.

This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded.

The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA).

APPLICABLE..

Except for primary-to-secondary leakage, the safety analyses SAFETY ANALYSES do not address operational leakage.

However, other operational leakage is.related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event. The safety analysis for an event resulting in steam discharge to.the atmosphere assumes a 1 gpm primary to secondary leakage as the. initial condition.

SEQUOYAH - UNIT 2-

.B'3/4 4-4e Amendment No. 211, 213,

221,

REACTOR COOLANT SYSTEM BASES Primary to secondary leakage is a factor in the dose releases outside containment resulting from a steam line break (SLB) accident.

To a lesser extent, other accidents i

or transients involve secondcry steam release to the atmosphere, such as a steam generator tube rupture (SGTR).

The leakage contaminates the secondary fluid.

The FSAR (Ref. 3) analysis for SGTR assumes the contaminated secondary fluid is released via safety valves for up to 30 minutes. Operator action is taken to isolate the affected steam generator within this time period. The 1 gpm primary to secondary leakage is relatively inconsequential.

The SLB is more limiting for site radiation releases.

The safety analysis for the SLB accident assumes 1 gpm primary to secondary leakage in one generator as an initial condition. The dose consequences resulting from the SLB accident are well within the limits defined in 10 CFR 100 or the staff approved licensing basis (i.e.,

a small fraction of these limits).

Based on the NDE uncertainties, bobbin coil voltage distribution and crack growth rate from the previous inspection, the expected leak rate following a steam line rupture is limited to below 8.21 gpm at atmospheric conditions and 70*F in the faulted loop, which R227 i

will limit the calculated offsite doses to within 10 percent

)

of the 10 CFR 100 guidelines.

If the projected and cycle J

distribution of crack indications results in primary-to-secondary leakage greater than 8.21 gpm in the faulted loop during a postulated steam line break event, additional tubes must be removed from service in order to reduce the postulated primary-to-secondary steam line break leakage to below 8.21 gpm.

The RCS operational leakage satisfies Criterion 2 of the NRC Policy Statement.

LCO RCS operational leakage shall be limited to:

a.

PRESSURE BOUNDARY LEAKAGE No PRESSURE BOUNDARY LEAKAGE is allowed, being indicative of material deterioration.

Leakage of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher leakage.

Violation of this LCO could result in continued degradation of the RCPB.

Leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

b.

UNIDENTIFIED LEAKAGE 1

One gpm of UNIDENTIFIED LEAKAGE is allowed as a I

reasonable minimum detectable amount that the i

containment air monitoring and containment pocket SEQUOYAH - UNIT 2 B 3/4 4-4f Amendment No. 211, 213,

227,

l

}

REACTOR COOLANT SYSTEM BASES sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO could result in continued degradation of the RCPB, if the leakage is'from the pressure boundary.

c.

Primary to Secondary Leakage through Any One Steam Generator (SG)

The 150 gallons per day limit on one SG is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line rupture.

If leaked through many cracks, the cracks are very small, and the above assumption is conservative.

The 150-gallons per day limit incorporated into Surveillance 4.4.6.2.1 is more restrictive than the standard operating leakage limit and is intended to provide an additional margin to accommodate a crack which might grow at a greater than expected rate or R227 unexpectedly extend outside the thickness of the tube support plate.

Hence, the reduced leakage limit, when combined with an effective leak rate monitoring program, provides additional assurance that, should a significant leak be experienced, it will be detected, and the plant shut down in a timely manner.

d.

IDENTIFIED LEAKAGE Up to 10 gpm of IDENTIFIED LEAKAGE is considered allowable because leakage is from known sources that do not interfere with detection of UNIDENTIFIED LEAKAGE and is well-within the capability of the RCS Makeup System.

IDENTIFIED LEAKAGE includes leakage to the containment from specifically known and located sources, but does not include PRESSURE BOUNDARY LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered leakage).

Violation of this LCO could result in continued degradation of a component or system.

APPLICABILITY In MODES 1, 2, 3,

and 4, the potential for reactor coolant PRESSURE BOUNDARY LEAKAGE is greatest when the RCS is pressurized.

In MODES 5 and 6, leakuge limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for leakage.

SEQUOYAH - UNIT 2 B 3/4 4-4g Amendment No. 211, 213,

227,

]

v-REACTOR ~ COOLANT SYSTEM BASES-LCO 3/4.4.6.3, "RCS Pressure Isolation Valve (PIV) Leakage,"

measures leakage through each individual PIV and can impact this LCO.

Of the two PIVs in series in each isolated line, leakage ~ measured through one PIV does not result in RCS leakage when the other is leak tight.

If both valves leak and result in'a loss of mass from the RCS, the loss must be included in the allowable. IDENTIFIED LEAKAGE.

ACTIONS Action a:

(_

If any PRESSURE BOUNDARY LEAKAGE exists, the reactor must be brought to lower pressure conditions to reduce the severity of the leakage and its potential consequences.

It should be noted that leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

The reactor must be brought to MODE 3 l

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

l This action reduces the leakage and also reduces the factors that tend to degrade _the pressure boundary.

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.

In MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration L

is much less likely.

Action b:

UNIDENTIFIED LEAKAGE, IDENTIFIED LEAKAGE, or primary-to-secondary leakage in excess of the LCO limits must be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

This completion time allows time to verify leakage rates and either identify UNIDENTIFIED LEAKAGE or reduce leakage to within limits before the reactor must be shut down.

This action is necessary to prevent further deterioration of the RCPB.

If UNIDENTIFIED LEAKAGE, IDENTIFIED LEAKAGE, or primary to secondary leakage cannot be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the reactor must be brought to lower pressure conditions to reduce the severity of the leakage and its potential consequences.

The reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

This action reduces the leakage and also reduces the factors that tend to degrade the pressure boundary.

The allowed completion times are reasonable, based on j

l operating experience, to reach the required plant conditions j

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

In MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.

l SEQUOYAH - UNIT 2 B 3/4 4-4h Amendment No. 211, 213, 1

227, 1

l

e i

REACTOR COOLANT SYSTEM BASES SURVEILLANCE Surveillance 4.4.6.2.1-REQUIREMENTS l

Verifying RCS leakage to be within the LCO limits ensures the integrity of the RCPB is maintained.

PRESSURE BOUNDARY L

LEAKAGE would at first appear.as UNIDENTIFIED LEAKAGE and I

can only be positively identified by inspection.

It should i

be noted that leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

UNIDENTIFIED LEAKAGE and IDENTIFIED LEAKAGE are determined by performance of an RCS water i

inventory balance.

Primary-to-secondary leakage is also measured by performance of an RCS water inventory, balance in conjunction with effluent monitoring within the secondary steam and feedwater systems.

The RCS water inventory' balance must be met with the reactor at steady state operating conditions (stable pressure, j

L temperature, power level, pressurizer and makeup tank I

levels, makeup, letdown, and RCP seal injection and return l

flows).

Therefore, a footnote is added allowing that this SR is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady state operation. The 12-hour allowance j

provides sufficient time to collect and process all necessary data after stable plant conditions are established.

Performance of.this surveillance within the

[

12-hour allowance is required to maintain compliance with the provisions of Specification 4.0.3.

l Steady state operation is required to perform a proper i

l inventory balance since calculations during maneuvering are not useful.

For RCS operational leakage determination by l

water inventory balance, steady' state is defined as stable t

.RCS pressure, temperature, power level, pressurizer and

[-

makeup tank levels, makeup and letdown, and RCP seal l

injection and return flows.

An early warning of PRESSURE BOUNDARY LEAKAGE or UNIDENTIFIED LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment' pocket sump level.

It should be noted that i

leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE. These leakage detection systems are specified in LCO 3/4.4 6.1,

" Leakage Detection Instrumentation."

j-The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months frequency is a reasonable interval to trend l-leakage and recognizes the importance of early leakage l

detection in the prevention of accidents.

Surveillance 4.4.6.2.2

This surveillance provides the means necessary'to determine SG OPERABILITY in an operational MODE.

The requirement to l

demonstrate SG tube integrity in accordance with the Steam Generator Tube Surveillance Program emphasizes the importance of SG tube integrity, even though this surveillance cannot be performed at normal operating

-conditions.

SEQUOYAH.- UNIT 2 B 3/4 4-41 Amendment No. 211, 213,

227,

REACTOR COOLANT SYSTEM BASES

................................o...........................................

l REFERENCES-1.

I 2.

Regulatory Guide 1.45, May 1973.

1 3.

FSAR, Section-15.4.3.

l J

i l

SEQUOYAH, UNIT 2 B 3/4 4-4j Amendment No. 211, 213,

227,

l l

REACTOR COOLANT SYSTCM BASES 3/4.4.6.3 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVE LEAKAGE BACKGROUND 10 CFR 50.2, 10 CFR 50.55a(c), and GDC 55 of i

10 CFR 50, Appendix A (Refs.

1, 2,.and 3), define reactor coolant system.(RCS) pressure isolation valves (PIVs) as any two normally closed valves in series within the reactor coolant pressure boundary (RCPS),

which separate the high pressure RCS from an attached low pressure system.

During their lives, these valves can produce. varying amounts of reactor coolant leakage through either normal operational wear or mechanical deterioration. The RCS PIV leakage LCO allows RCS high pressure operation when leakage through these valves exists in amounts that do not compromise safety.

Although this specification provides a limit on allowable PIV leakage rate, its main purpose is to prevent overpressure failure of the low pressure portions of connecting systems.

The leakage limit is an indication that the PIVs between the RCS and the connecting systems are degraded or degrading.

PIV leakage could lead to overpressure of the low pressure r

I piping or components.

Failure consequences could be a loss of coolant accident (LOCA) outside of containment or an unanalyzed accident that could degrade the ability for low pressure injection.

L The basis for this LCO is the 1975 NRC, " Reactor

(

Safety Study," (Ref. 4) that identified potential l

intersystem LOCAs as a significant contributor to the risk of core melt.

.A subsequent study (Ref. 5) evaluated various PIV configurations to deterinine the probability of intersystem LOCAs.

PIVs are provided to isolate the RCS from the following typically connected systems:

a.

Residual Heat Removal (RHR) System; b.

Safety Injection System; and c.

Chemical and Volume Control System.

The PIVs are listed in Table 3.4-1.

Violation of this LCO could result in continued degradation of a PIV, which could lead to overpressurization of a low pressure system and the loss of the integrity of a fission product barrier.

- SEQUOYAH - UNIT 2 B 3/4 4-4k Amendment No.

m

r i

1 REACTOR COOLANT SYSTEM BASES APPLICABLE Reference 4 identified potential intersystem LOCAs as SAFETY ANALYSES a significant contributor to the risk of core melt.

The dominant accident sequence in the intersystem LOCA category is the failure of the low pressure portion of the RHR System outside of containment.

The accident is the result of a postulated failure of the PIVs, which are part of the RCPB, and the subsequent pressurization of tha RHR System downstream of the PIVs from the RCS.

Because the low pressure portion of the RHR System is typically designed for 600 psig, j

overpressurization failure of the RHR low pressure line would result in a LOCA outside containment and subsequent risk of core melt.

3 Reference 5 evaluated various PIV configurations, leakage testing of the valves, and operational changes to determine the effect on the probability of intersystem LOCAs.

This study concluded that periodic leakage testing of the PIVs can substantially reduce the probability of an intersystem LOCA.

RCS PIV leakage satisfies Criterion 2 of the NRC Policy Statement.

I LCO RCS PIV leakage is IDENTIFIED LEAKAGE into closed systems connected to the RCS.

Isolation valve leakage is usually on the order of drops per minute.

Leakage that increases significantly suggests that something is operationally wrong and corrective action must be taken.

The LCO PIV leakage limit is 0.5 gpm per nominal inch I

of valve size with a maximum limit of 5 gpm.

The l

previous criterion of 1 gpm for all valvo sizes imposed an unjustified penalty on the larger valves without providing information on potential valve degradation and resulted in higher personnel radiation i

exposures.

A study concluded a leakage rate limit j

based on valve size was superior to a single allowable i

value.

i Reference 6 permits leakage testing at a lower j

pressure differential than between the specified

~

maximum RCS pressure and the normal pressure of the connected system during RCS operation (the maximum pressure differential) in those types of valves in which the higher service pressure will tend to diminish the overall leakage channel opening.

In such cases, the observed rate may be adjusted to the maximum pressure differential by assuming leakage is directly proportional to the pressure differential to i

(

the one half power.

1 SEQUOYAH - UNIT 2 B 3/4 4-41 Amendment No.

L

REACTOR COOLANT SYSTEM BASES APPLICABILITY In MODES 1, 2, 3, and 4, this LCO applies because the PIV leakage potential is greatest when the RCS is pressurized.

In MODE 4, valves in the RHR flow path are not required to meet the requirements of this LCO when in, or during the transition to or from, the RHR mode of operation.

In MODES 5 and 6, leakage limits are not provided because the lower reactor coolant pressure results in a reduced potential for leakage'and for a LOCA outside the containment.

ACTIONS Action a:

)

The flow path must be isolated. Action a is modified by a note that the valves used for isolation must meet the same leakage requirements as the PIVs and must be within the RCPB.

Action a requires that the isolation with one valve must be performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Four hours provides time to reduce leakage in excess of the allowable limit and to isolate the affected system if leakage cannot be reduced.

The 4-hour completion time allows the actions and restricts the operation with leaking isolation valves.

The 72-hour completion time after exceeding the limit allows for the restoration of the leaking PIV to OPERABLE status.

This timeframe considers the time required to complete this action and the low probability of a second valve failing during this period.

If leakage cannot be reduced or the system isolated, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . 'This Action may reduce the leakage and also reduces the potential for a LOCA outside the containment.

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.

Action b:

Action b provides clarification that each flow path allows separate entry into Action a This is allowed based upon the functional independence of the flow path.

SEQUOYAH - UNIT 2 B 3/4 4-4m

' Amendment No.

REACTOR COOLANT SYSTEM

BASES, Action c:

Action ( requires an evaluation of affected systems if a PIV is inoperable.

The leakage may have affected system operability or isolation of a leaking flow path with an alternate valve may have degraded the ability of.the interconnected system to perform its' safety function.

SURVEILLANCE Surveillance 4.4.6.3 REQUIREMENTS Performance of leakage testing on each RCS PIV~or isolation valve used to satisfy Action a is required to verify that leakage is below the specified limit and to identify each leaking. valve.

The leakage limit of 0.5 gpm per inch of nominal valve diameter up to 5 gpm maximum applies to each valve.

Leakage testing requires a stable pressure condition, q

For the two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves.

If the PIVs are l

not individually leakage tested, one valve may have failed completely and not be detected if the other valve in series meets the leakage requirement.

In this situation, the protection provided by redundant valves would be lost.

Testing is to be performed every 18 months, a typical refueling cycle, if the plant does not go'into MODE 5 for at least 7 days.

The 18 month frequency is consistent with 10 CFR 50.55a(g) (Ref. 7) as contained in the Inservice Testing Program, is within frequency allowed by the American Society of Mechanical Engineers (ASME) Code,Section XI (Ref. 6),'and is based on the need to perform such surveillances under the conditions that apply during an outage'and the potential for an unplanned transient if the surveillances were performed with the reactor at power.

In addition, testing must be performed once after the valve has been opened by flow or exercised to ensure tight reseating.

PIVs disturbed in the performance of this surveillance should also be tested unless documentation shows that an infinite testing loop

.cannot practically be avoided.

Testing must be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the valve has been reseated. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a reasonable and

practical time limit for performing this test after opening or. reseating a valve.

SEQUOYAH - UNIT 2 B 3/4 4-4n Amendment No.

s.

\\

REACTOR COOLANT SYSTEM BASES The leakage limit is to be met at the RCS pressure associated with MODES 1 and 2.

This permits leakage testing at high differential pressures with stable conditions not possible in the MODES with lower pressures.

Entry into MODES 3 and 4 is allowed to establish the necessary differential pressures and stable conditions to allow for performance of this surveillance.

The note that allows this provision is complementary to the frequency of prior to entry into MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months.

In addition, this surveillance is not required to be performed on the RHR System when the RHR System is aligned to the RCS in the shutdown cooling mode of operation.

PIVs contained in the RHR shutdown cooling flow path must be leakage rate tested after RHR is secured and stable unit conditions and the necessary differential pressures are established.

REFERENCES 1.

2.

10 CFR 50.55a(c).

3.

10 CFR 50, Appendix A, Section V, GDC 55.

4.

WASH-1400 (NUREG-75/014), Appendix V, October 1975.

5.

NUREG-0677, May 1980.

6.

ASME, Boiler and Pressure vessel Code,Section XI.

7.

i 1

l SEQUOYAH - UNIT 2 B 3/4 4-4o Amendment No.

)

EMERGENCY CORE COOLING SYSTEM BASES I

3/4.5.6 SEAL INJECTION FLOW BACKGROUND The function of the seal injection throttle valves during an accident is similar to the function of the ECCS throttle valves in that each restricts flow from the centrifugal charging pump header to the Reactor Coolant System (RCS).

j J

The restriction on reactor coolant pump (RCP) seal injection flow limits the amount of ECCS flow that would be diverted from the injection path following an accident.

This limit is based on safety analysis assumptions that are required because RCP seal injection flow is not isolated during safety injection.

APPLICABLE All ECCS subsystems are taken credit for in the large SAFETY ANALYSES break loss of coolant accident (LOCA) at full power (Ref. 1).

The LOCA analysis establishes the minimum flow for the ECCS pumps.

The centrifugal charging pumps are also credited in the small break LOCA analysis.

This analysis establishes the flow and j

discharge head at the design point for the centrifugal i

charging pumps.

The steam generator tube rupture and main steam line break event analyses also credit the centrifugal charging pumps, but are not limiting in their design.

Reference to these analyses is made in assessing changes to the Seal Injection System for evaluation of their effects in relation to the acceptance limits in these analyses.

This LCO ensures that seal injection flow will be sufficient for RCP seal integrity but limited so that the ECCS trains will be capable of delivering sufficient water to match boiloff rates soon enough to minimize uncovering of the core following a large LOCA.

It also ensures that the centrifugal charging pumps will deliver sufficient water for a small LOCA and sufficient boron to maintain the core suberitical.

For smaller LOCAs, the charging pumps alone deliver sufficient fluid to overcome the loss and maintain RCS inventory.

Seal injection flow satisfies Criterion 2 of the NRC Policy Statement.

SEQUOYAH - UNIT 2 B 3/4 5-4 Amendment No.

EMERGENCY CORE COOLING SYSTEM BASES LCO The intent of the LCO limit on seal injection flow is to make sure that flow through the RCP seal water injection line is low enough to ensure that sufficient centrifugal charging pump injection. flow is directed-

~

to the RCS via the injection points (Ref. 2).

The LCO is not strictly a flow limit, but rather a flow limit based on a flow line resistance.

In order to establish the proper flow line resistance, a pressure and flow must be known.. The flow line resistance is established by adjusting the RCP seal injection needle valves to provide a total seal injection flow in the acceptable region of Figure B 3.5.6-1.

The centrifugal charging pump discharge header pressure remains essentially constant through all the applicable MODES of this LCO.

A reduction in RCS pressure would result in more flow being diverted to the RCP seal injection line than at normal operating pressure.

The valve settings established at the prescribed centrifugal charging pump discharge header pressure result in a conservative valve position should RCS pressure decrease. The flow limite established by Figure B 3.5.6-1 are consistent with the accident analysis.

The limits on seal injection flow must be met to render the ECCS OPERABLE.

If these conditions are not met, the ECCS flow will not be as assumed in the accident analyses.

APPLICABILITY In MODES 1, 2, and 3, the seal injection flow limit is dictated by ECCS flow requirements, which are specified for MODES 1, 2,

3, and 4.

The seal injection flow limit is not applicable for. MODE 4 and lower, however, because high seal injection flow is less critical as a result of-the lower initial RCS pressure and decay heat removal requirements in these MODES. Therefore, RCP seal injection flow must be limited in MODES 1, 2, and 3 to ensure adequate ECCS performance.

SEQUOYAH - UNIT 2-B 3/4 5-5 Amendment No.

l l

EMERGENCY CORE COOLING SYSTEM BASES ACTION With the seal injection flow exceeding its limit, the amount of charging flow available to the RCS may be reduced.

Under this condition, action must be taken to restore the flow to below its limit.

The operater

)

has 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from the time the flow is known to b 1

above the limit to correctly position the manual valves and thus be in compliance with the accident analysis.

The completion time minimizes the potential exposure of the plant to a LOCA with insufficient injection flow and provides a reasonable time to restore seal injection flow within limits.

This time is conservative with respect'to the completion times of other ECCS LCOs; it is based on operating experience and is sufficient for taking corrective actions by operations personnel.

When the actions cannot be completed within the required completion time, a controlled shutdown must be initiated. The completion time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for reaching MODE 3 from MODE 1 is a reasonable time for a l

controlled shutdown, based on operating experience and normal cooldown rates, and does not challenge plant safety systems or operators.

Continuing the plant shutdown from MODE 3, an additional-6 hours is a reasonable time, based on operating experience and normal cooldown rates, to reach MODE 4, where this LCO is no longer applicable.

1 SURVEILLANCE Surveillance 4.5.6 REQUIREMENTS Verification every 31 days that the manual seal injection throttle valves are adjusted to give a flow within the limit ensures that proper manual seal injection throttle valve position, and hence, proper seal injection flow, is maintained.

The differential l

pressure taat is above the reference minimum value is established between the charging header (PT 62-92) and the RCS, and total seal injection flow is verified to be within the limits determined in accordance with the i

ECCS safety analysis (Ref. 3).

The seal water injection flow limits are shown in Figure B 3.5.6-1.

l.

The frequency of 31 days is based on engineering judgment and is consistent with other ECCS valve surveillance _ frequencies. The frequency has proven to be acceptable through operating experience.

The requirements for charging flow vary widely according to plant status and configuration.

When charging flow is adjusted, the positions of the air-operated valves, which control charging flow, are SEQUOYAH - UNIT 2 B 3/4 5-6 Amendment No.

L_

I l

1

\\

l EMERGENCY CORE COOLING SYSTEM BASES adjusted to balance the flows through the charging header and through the seal injection header to ensure that the seal injection flow to the RCPs is maintained between 8 and 13 gpm per pump.

The reference minimum differential pressure across the seal injection needle valves ensures that regardless of the varied settings of the charging flow control valves that are required to support optimum charging flow, a reference test I

condition can be established to ensure that flows l

across the needle valves are within the safety analysis.

The values in the safety analysis for this reference set of conditions are calculated based on conditions during power operation and they are correlated to the minimum ECCS flow to be maintained under the most limiting accident conditions.

As noted, the surveillance is not required to be l

performed until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the RCS pressure has l

stabilized within a i 20 psig range of normal l

operating pressure.

The RCS pressure requirement is specified since this configuration will produce the required pressure conditions necessary to assure that i

the manual valves are set correctly.

The exception is l

limited to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to ensure that the surveillance is timely.

Performance of this surveillance within the l

4-hour allowance is required to maintain compliance i

l with the provisions of Specification 4.0.3.

l l

REFERENCES 1.

FSAR, Chapter 6.3 " Emergency Core Cooling System" and Chapter 15.0 " Accident Analysis".

1 l

2.