ML17157C446

Amends 129 & 98 to Licenses NPF-14 & NPF-22,respectively, Changing TSs to Decrease Test Frequency of drywell-to- Suppression Chamber Bypass Test to Coincide W/Test Frequency for Primary Containment Integrated Leak Rate Test

ML17157C446
Person / Time
Site:	Susquehanna
Issue date:	08/11/1993
From:	Boyle M Office of Nuclear Reactor Regulation
To:
Shared Package
ML17157C447	List: ML17157C446 ML17157C448
References
NUDOCS 9308300308
Download: ML17157C446 (23)

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 PENNSYLVANIA POWER

& LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.

DOCKET NO. 50-387 SUS UEHANNA STEAM ELECTRIC STATION UNIT 1

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 129 License No. NPF-14 1.

The Nuclear Regulatory Commission (the Commission or the NRC) having found that:

A.

The application for the amendment filed by the Pennsylvania Power Light Company, dated May 4,

1993, and supplemented by letter dated July 15,

1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; C.

There is reasonable assurance:

(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I;.

D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

9308300308 9308ii PDR ADOCL 05000387(

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Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No.

NPF-14 is hereby amended to read as follows:

(2)

Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.

129 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

PP&L shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

August ll,:1993 Michael L.

yle, Acting Director Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

TTACHMENT TO LICENSE AMENDMENT NO.

129 C'ACILITY OPERATING LICENSE NO. NPF-14 DOCKET NO. 50-387 Replace the following pages of the Appendix A Technical Specifications with enclosed pages.

The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

The overleaf pages are provided to maintain document completeness.*

REMOVE 3/4 6-13 3/4 6-14 3/4 6-3 3/4 6-4 B 3/4 6-5 INSERT 3/4 6-13*

3/4 6-14 3/4 6-3*

3/4 6-4 B 3/4 6-5

t CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION Continued

~

'CTION:

C.

d.

e.

(Continued) b)

110 F, place the reactor mode switch in the Shutdown posi-tion and operate at least one residual heat removal loop in the suppression pool cooling mode.

3.

Wi~ the suppression chamber average water temperature greater than 120 F, depressurize the reactor pressure vessel to less than 200 psig within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

With only one suppression chamber water level indicator OPERABLE and/or with less than eight suppression pool water temperature indicators covering at least six locations OPERABLE, restore the inoperable indicator(s) to OPERABLE status within 7 days or verify suppression chamber water level and/or temperature to be within the limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

With no suppression chamber water level indicators OPERABLE and/or with less than one suppression pool water temperature indicator at at least six different locations OPERABLE, restore at least one water level indicator and at least one water temperature indicator at at least six different locations to OPERABLE status within 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> s or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

With the drywell-to-suppression chamber bypass leakage in excess of the limit, restore the bypass leakage to within the limit prior to increasing"reactor coolant temperature above 2004F.

SURVEILLANCE RE UIREMENTS 4.6.2.1 a ~

b.

The suppression chamber shall be demonstrated OPERABLE:

By verifying the suppression chamber water volume to be within the limits at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in OPERATIONAL CONDITION 1 or 2 by verifying the suppression chamber average water temperature to be less than or equal to 904F, except:

1.

At least once per 5 minutes during testing which adds heat to the suppression

chamber, by verifying the suppression chamber average water temperature less than or equal to 105 F.

2.

At least once per hour when suppression chamber average water temperature is greater than or equal to 904F, by verifying:

a)

Suppression chamber average water temperature to be less than or equal to 1104F, and b)

THEP.'.",.',

POWER to be less than or equal to lX of RATED THERY<<'iL POWER after suppression chamber average water temperature has exceeded 904F for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.

At least once per 30 minutes following a scram with suppression chamber average water temperature greater than or equal to 904F, by verifying suppression chamber average water temperature less than or equal to 1204F.

SUSQUEHANNA - UNIT 1 3/4 6-13 Amendment No.

82 AgG 3 0 1988

CONTAINMENTSYSTEMS SURVEILLANCE REQUIREIVIENTS Continued c.

By verifying at least two suppression chamber water level indicators and at least sixteen surface water temperature indicators, at least one pair in each suppression pool sector, OPERABLE by performance of a:

1.

CHANNEL CHECK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 2.

CHANNEL FUNCTIONALTEST at least once per 31 days, and 3.

CHANNEL CALIBRATIONat least once per 18 months, with the water level and temperature alarm setpoint for:

1.

High water level 6 23'9",

2.

Low water level R 22'3", and 3.

High water temperature:

a)

First setpoint, 6 90'F, b)

Second setpoint, 6 105'F, c)

Third setpoint, M 110 F, and d)

Fourth setpoint, 6 120'F.

d.

By conducting a dryweii-to-suppression chamber bypass leak test at an initial differential pressure of at least 4.3 psi and verifying that the AP/k calculated from the measured leakage is within the specified limit. The bypass leak test shall be conducted at 40 ~ 10 month intervals during shutdown, during each 10 year service period. Ifany drywell-to-suppression chamber bypass leak test fails to meet the specified limit, the test schedule for subsequent tests shall be reviewed and approved by the Commission.

Iftwo consecutive tests fail to meet the specified limit, a test shall be performed at least every 18 months until two consecutive tests meet the specified limit, at which time the above test schedule may be resumed.

e.

By conducting a leakage test on the drywell-to-suppression chamber vacuum breakers at a differential pressure of at least 4.3 psi and verifying that the total leakage area A/(k)"~ contributed by all vacuum breakers is less than or equal to 30% of the specified limit and the leakage area for an individual set of vacuum breakers is less than or equal to 12% of the specified limit. The vacuum breaker leakage test shall be conducted during each refueling outage for which the drywell-to-suppression chamber bypass leak test in Specification 4.6.2.1.d is not conductey'.

SUSQUEHANNA - UNIT 1 3/4 6-14 Amendment No. ~29 129

CONTAINMENTSYSTEMS SURVEILLANCE REQUIREMENTS Co tinued)

I c.

By verifying at least two suppression chamber water level indicators and at least sixteen surface water temperature indicators, at least one pair in each suppression pool sector, OPERABLE by performance of a:

1.

CHANNEL CHECK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 2.

CHANNEL FUNCTIONALTEST at least once per 31 days, and 3.

CHANNEL CALIBRATIONat least once per 18 months, with the water level and temperature alarm setpoint for:

1.

High water level s 23'9",

2.

Low water level > 22'3", and 3.

High water temperature:

a)

First setpoint, 6 90'F, b)

Second setpoint, C 105'F, c)

Third setpoint, c 110'F, and d)

Fourth setpoint, 6 120'F.

d.

By conducting a dryweii-to-suppression chamber bypass leak test at an initial differential pressure of at least 4.3 psi and verifying that the A)V k calculated from the measured leakage is within the specified limit. The bypass leak test shall be conducted at 40 a 10 month intervals during shutdown, during each 10 year service period.

If any drywall-to-suppression chamber bypass leak test fails to meet the specified limit, the test schedule for subsequent tests shall be reviewed and approved by the Commission.

Iftwo consecutive tests fail to meet the specified limit, a test shall be performed at least every 18 months until two consecutive tests meet the specified limit, at which time the above test schedule may be resumed.

e.

By conducting a leakage test on the drywell-to-suppression chamber vacuum breakers at a differential pressure of at least 4.3 psi and verifying that the total leakage area A/(k)"~ contributed by all vacuum breakers is less than or equal to 30% of the specified limit and the leakage area for an individual set of vacuum breakers is less than or equal to 12% of the specified limit. The vacuum breaker leakage test shall be conducted during each refueling outage for which the drywell-to-suppression chamber bypass leak test in Specification 4.6.2.1.d is not conducted.

SUSQUEHANNA - UNIT 1 3/4 6-14 Amendment No. ~29 129

CONTAINMENT SYSTEMS BASES 3/4. 6. 2 OEPRESSURIZATION SYSTEMS The specifications of this section ensure that the primary containment pressure will not exceed the design pressure of 53 psig during primary system blowdown from full operating pressure.

The suppression chamber water provides the heat sink for the reactor coolant system energy release following a postulated rupture of the system.

The suppres-sion chamber water volume must absorb the associated decay and structural sensible heat released during reactor coolant system blowdown from 1055 psig.

Since all of the gases in the drywell are purged into the suppression chamber air space during a loss of coolant accident, the pressure of the liquid must not exceed 53 psig, the suppression chamber maximum pressure.

The design volume of the suppression

chamber, water and air, was obtained by considering that the total volume of reactor coolant and to be considered is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.

Using the minimum or maximum water volumes given in thfs specification, containment pressure during the design basis accident is approximately 45.0 psig which is b~low the design pressure of 53 psig.

Maximum water volume of 133,540 ft results in a downcome'r submergence of 12 feet and the minimum volume of 122,410 ft results in a submergence approximately 24 inches less.

The majority of the Bodega tests were run'with a submerged length of four feet and with complete condensation.

Thus, with respect to the downcomer submergence, this specification is adequate.

The maximum temperature at the end of the blow-down tested during the Humboldt Bay and Bodega.,gay tests was 170 F and this is conservatively taken to be the limit for complete condensation of the reactor

coolant, although condensation would occur for temperatures above 1704F.

Should it be necessary to make the suppression chamber inoperable, this shall only be done as specified in Specification 3.5.3.

Under full power operating conditions, blowdown from an initial suppression chamber water temperature of 90oF results in a water temperature of approx-imately 1284F imaediately following blowdown which is below the 1704F used for complete condensation via T-quencher devices.

At this temperature and atmos-pheric pressure, the available NPSH exceeds that required by both the RHR and core spray pumps, thus there is no dependency on containment overpressure dur ing the accident injection phase.

If both RHR loops are used for containment cooling, there is no dependency on containment overpressure for post-LOCA operations.

Experimental data indic';tes th't excessive steam condensing loads can be avoided if the peak local teml:eratur1 of the suppression pool is maintained below 200 F during any period of relief valve operation.

Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings.

SUS(UEHANNA - UNIT 1 8 3/4 6-3 Amendment No.

29

CONTAINMENTSYSTEMS BASES DEPRESSURIZATION SYSTEMS Continued)

Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends.

By requiring the suppression pool temperature to be frequently recorded during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken.

The requirement for an external visual examination following any event where potentially high loadings could occur provides assurance that no significant damage was encountered.

Particular attention should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to be the points of highest stress.

In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a safety-relief valve inadvertently opens or sticks open.

As a minimum this action shall include:

(1) use of all available means to close the valve, (2) initiate suppression pool water cooling, (3) initiate reactor shutdown, and (4) if other safety-relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open safety relief valve to assure mixing and uniformity of energy insertion to the pool.

During a LOCA, potential leak paths between the drywall and suppression chamber airspace could result in excessive containment pressures, since the steam flow into the airspace would bypass the heat sink capabilities of the pool.

Potential sources of bypass leakage are the suppression chamber-to-drywell vacuum breakers (VBs), penetrations in the diaphragm floor, and cracks in the diaphragm floor/liner plate and downcomers located in the suppression chamber airspace.

The containment pressure response to the postulated bypass leakage can be mitigated by manually actuating the suppression chamber sprays. An analysis was performed for a design bypass leakage area of A/(k)'~equal to 0.0535 ft~ to verify that the operator has sufficient time to initiate the sprays prior to exceeding the containment design pressure of 53 psig.

The limit of 10% of the design value of 0.0535 ft~ ensures that the design basis for the steam bypass analysis is met.

The drywall-to-suppression chamber bypass test at a differential pressure of at least 4.3 psi verifies the overall bypass leakage area for simulated LOCA conditions is less than the specified limit. For those outages where the drywell-to-suppression chamber bypass leakage test is not conducted, the VB leakage test verifies that the VB leakage area is less than the bypass limit, with a 70% margin to the bypass limit to accommodate the remaining potential leakage area through the passive structural co'mponents.

Previous drywall-to-suppression chamber bypass test data indicates that the bypass leakage through the passive structural components will be much less than the 70% margin.

The VB leakage limit, combined with the negligible passive structural leakage area, ensures that the drywell-to-suppression chamber bypass leakage limit is met for those outages for which the drywell-to-suppression chamber bypass test is not scheduled.

3/4.6.3 PRIMARY CONT/'.IK51CN'!>,.OLi'.T'f'P HALVES The OPERABILITY of the prima;y containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment and is consistent with the requirements of GDC 54 through 57 of Appendix A to 10CFR 50.

Containment isolation within the time limits specified for those isolation valves designed to close automatically SUSQUEHANNA - UNIT 1 B 3/4 6-4 Amendment No. 129

CONTAINMENTSYSTEMS BASES DEPRESSURIZATION SYSTEMS Continued)

Because of the large, volume and thermal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends.

By requiring the suppression pool temperature to be frequently recorded during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken.

The requirement for an external visual examination following any event where potentially high loadings could occur provides assurance that no significant damage was encountered.

Particular attention should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to be the points of highest stress.

In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a safety-relief valve inadvertently opens or sticks open.

As a minimum this action shall include:

(1) use of all available means to close the valve, (2) initiate suppression pool water cooling, (3) initiate reactor shutdown, and (4) if other safety-relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open safety relief valve to assure mixing and uniformity of energy insertion to the pool.

Ouring a LOCA, potential leak paths between the drywell and suppression chamber airspace could result in excessive containment pressures, since the steam flow into the airspace would bypass the heat sink capabilities of the pool.

Potential sources of bypass leakage are the suppression chamber-to-drywell vacuum breakers (VBs), penetrations in the diaphragm floor, and cracks in the diaphragm floor/liner plate and downcomers located in the suppression chamber airspace.

The containment pressure response to the postulated bypass leakage can be mitigated by manually actuating the suppression chamber sprays.

An analysis was performed for a design bypass leakage area of A/(k)'"equal to 0.0535 ft'o verify that the operator has sufficient time to initiate the sprays prior to exceeding the containment design pressure of 53 psig.

The limit of 10% of the design value of 0,0535 ft~ ensures that the design basis for the steam bypass analysis is met, The drywell-to-suppression chamber bypass test at a differential pressure of at least 4.3 psi verifies the overall bypass leakage area for simulated LOCA conditions is less than the specified limit. For those outages where the drywall-to-suppression chamber bypass leakage test is not conducted, the VB leakage test verifies that the VB leakage area is less than the bypass limit, with a 70% margin to the bypass limit to accommodate the remaining potential leakage area through the passive structural components.

Previous drywell-to-suppression chamber bypass test data indicates that the bypass leakage through the passive structural components will be much less than the 70% margin.

The VB leakage limit, combined with the negligible passive structural leakage area, ensures that the drywell-to-suppression chamber bypass leakage limit is met for those outages for which the drywell-to-suppression chamber bypass test is not scheduled.

3/4.6.3 PRIMARY CONT/'.INMENTISOLATIONVALVES, The OPERABILITY of the primary containment isolation valy'es. ensures that the containment atmosphere willbe isolated from the outside environment in-]he event of a release of radioactive material to the containment atmosphere or pressurization of the containment and is consistent with the requirements of GDC 54 through 57.of Appendix A to 10CFR 50, Containment isolation within the time limits specified for those isolation valves designed to close automatically SUSQUEHANNA - UNIT 1 B 3/4 6-4 Amendment No. 129

CONTAINMENTSYSTEMS BASES ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.

3 4.6.4 VACUUM RELIEF Vacuum relief breakers are provided to equalize the pressure between the suppression chamber and drywell. This system willmaintain the structural integrity of the primary containment under conditions of large differential pressures.

The vacuum breakers between the suppression chamber and the drywall must not be inoperable in the open position since this would allow bypassing of the suppression pool in case of an accident.

There are five pairs of valves to provide redundancy so that operation may continue for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with no more than one pair of vacuum breakers inoperable in the closed position.

3 4.6.5 SECONDARY CONTAINMENT Secondary containment is designed to minimize any ground level release of radioactive material which may result from an accident.

The Reactor Building provides secondary containment during normal operation when the drywell is sealed and in service.

When the reactor is in COLD SHUTDOWN or REFUELING, the drywall may be open and the Reactor Building then becomes the only containment.

Establishing and maintaining a vacuum in the reactor building with the standby gas treatment system once per 18 months, along with the surveillance of the doors, hatches and dampers, is adequate to ensure that there are no violations of the integrity of the secondary containment.

The OPERABILITYof the standby gas treatment system ensures that sufficient iodine removal capability will be available in the event of a LOCA.

The reduction in containment iodine inventory reduces the resulting, site boundary radiation doses associated with containment leakage.

The operation of this system and resultant iodine removal capacity are consistent with the assumptions used in the LOCA analyses.

Cumulative operation of the system with the heaters OPERABLE for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> over a 31 day period is sufficient to reduce. the buildup of moisture on the adsorbers and HEPA filters.

F 3 4.6.6 PRIMARY CONTAINMENTATMOSPHERE CONTROL'.

The OPERABILITYof the systems required for the detection and control of hydrogen gas ensures that these systems will be available to maintain the hydrogen concentration within the primary containment below its flammable limit during post-LOCA conditions.

The drywell and suppression chamber hydrogen recombiner system is capable of controlling the expected hydrogen generation associated with (1) zirconium-water reactions, (2) radiolytic decomposition of water and (3) corrosion of metals within containment.

Drywell hydrogen mixing is provided to ensure adequate mixing of the containment atmosph re following a LOCA. This mixing action will prevent localized accumulations of hydrogen from exceeding the flammable limit.

The hydrogen control system is consistent with the recommendations of Regulatory Guide 1.7, "Control of Combustible Gas Concentrations in Containment Following a LOCA", March 1971.

SUSQUEHANNA - UNIT 1 B 3/4 6-5 Amendment No. 129

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 PENNSYLVANIA POWER

& LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.

DOCKET NO. 50-388 SUS UEHANNA STEAM ELECTRIC STATION UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 98 License No.

NPF-22 1.

The Nuclear Regulatory Commission (the Commission or the NRC) having found that:

A.

The application for the amendment filed by the Pennsylvania Power 8

Light Company, dated May 4,

1993, and supplemented by letter dated July 15, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; C.

There is reasonable assurance:

(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No.

NPF-22 is hereby amended to read as follows:

(2) Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.

98 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

PP&L shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3, This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

August 11, 1993 michael L. Boy A ting Director Project Directorate I-2 Division of Reactor Projects -'/II Office of Nuclear Reactor Regulation

ATTACHHENT TO LICENSE AHENDHENT NO.

ACILITY OPERATING LICENSE NO. NPF-22 DOCKET NO. 50-388 Replace the following pages of the Appendix A Technical Specifications with enclosed pages.

The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

The overleaf pages are provided to maintain document completeness.*

REHOVE 3/4 6-13 3/4 6-14 3/4 6-3 3/4 6-4 B 3/4 6-5 INSERT 3/4 6-13*

3/4 6-14 3/4 6-3*

3/4 6-4 B 3/4 6-5

CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION Continued ACTION:

(Continuedj C.

d.

e.

3.

With the suppression chamber average water temperature greater than 120 F, depressurize the reactor pressure vessel to less than 200 psig within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

With only one suppression chamber water level indicator OPERABLE and/or with less than eight suppression pool water temperature indicators covering at least six locations OPERABLE, restore the inoperable indicator(s) to OPERABLE status within 7 days or verify suppression chamber water level and/or temperature to be within the limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

With no suppression chamber water level indicators OPERABLE and/or with less than six suppression pool water temperature indicators covering at least six locations OPERABLE, restore at least one water level indicator and at least six water temperature indicators to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

With the drywell-to-suppression chamber bypass leakage in excess of the limit, restore the bypass leakage to within the limit prior to increasing reactor coolant temperature above 2004F.

SURVEILLANCE RE UIREMENTS 4.6.2.1 The suppression chamber shall be demonstrated OPERABLE:

a 0 b.

By verifying the suppression chamber water volume to be within the limits at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in OPERATIONAL CONDITION 1 or 2 by verifying the suppression chamber average water temperature to be less than or equal to 90 F, except:

1.

At least once per 5 minutes during testing which adds heat to the suppression

chamber, by verifying the suppression chamber average water temperature less than or equal to 105'F.

2.

At least once per hour when suppression chamber average water temperature is greater than or equal to 90'F, by verifying:

a)

Suppression chamber average water temperature to be less than or equal to 110 F, and b)

THERMAL POWER to be less than or equal to 1X of RATED THERMAL POWER after suppression cham86r average water temperature has exceeded 90 F for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.

At least once per 30 minutes following a scram with suppression chamber average water temperature greater than or equal to

904F, by verifying suppression chamber average water temperature less than or equal to 120 F.

SUS(UEHANNA - UNIT 2 3/4 6"13

CONTAINMENTSySTEMS SURVEILLANCE REQUIREIVIENTS Continued c.

By verifying at least two suppression chamber water level indicators and at least sixteen surface water temperature indicators, at least one pair in each suppression pool sector, OPERABLE by performance of a:

1.

CHANNEL CHECK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 2.

CHANNEL FUNCTIONAL TEST at least once per 31 days, and 3.

CHANNEL CALIBRATIONat least once per 18 months, with the water level and temperature alarm setpoint for:

1.

High water level 6 23'9",

2.

Low water level > 22'3", and 3.

High water temperature:

a)

First setpoint, 6 90'F, b)

Second setpoint, 6 105'F, c)

Third setpoint, 6 110'F, and d)

Fourth setpoint, ~ 120'F.

d.

By conducting a drywell-to-suppression chamber bypass leak test at an initial l

differential pressure of at least 4.3 psi and verifying that the A)V k calculated from the measured leakage is within the specified limit. The bypass leak test shall be conducted at 40 a 10 month intervals during shutdown, during each 10 year service period. Ifany drywall-to-suppression chamber bypass leak test fails to meet the specified limit, the test schedule for subsequent tests shall be reviewed and approved by the Commission.

Iftwo consecutive tests fail to meet the specified limit, a test shall be performed at least every 18 months until two consecutive tests meet the specified limit, at which time the above test schedule may be resumed.

e.

By conducting a leakage test on the drywell-to-suppression chamber vacuum breakers at a differential pressure of at least 4.3 psi and verifying that the total leakage area A/(k)" contributed by all vacuum breakers is less than or equal to 30% of the specified limit and the leakage area for an individual set of vacuum breakers is less than or equal to 12% of the specified limit. The vacuum breaker leakage test shall be conducted during each refueling outage for which the drywell-to-suppression chamber bypass leak test in Specification 4.6.2.1.d is not conducted.

SUSQUEHANNA - UNIT 2 3/4 6-14 Amendment No. 98

CONTAINMENTSYSTEMS SURVEILLANCE REQUIREMENTS Continued c.

By'verifying at least two suppression chamber water level indicators and at least sixteen surface water temperature indicators, at least one pair in each suppression pool sector, OPERABLE by performance of a:

1.

CHANNEL CHECK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 2.

CHANNEL FUNCTIONALTEST at least once per 31 days, and 3.

CHANNEL CALIBRATIONat least once per 18 months, with the water level and temperature alarm setpoint for:

1.

High water level 6 23'9",

2.

Low water level a 22'3", and 3.

High water temperature:

a)

First setpoint, ~ 90'F, b)

Second setpoint, 6 105'F, c)

Third setpoint, 6 110'F, and d)

Fourth setpoint, c 1204F.

d.

By conducting a drywell-to-suppression chamber bypass leak test at an initial l

differential pressure of at least 4.3 psi and verifying that the A/V k calculated from the measured leakage is within the specified limit. The bypass leak test shall be conducted at 40 a 10 month intervals during shutdown, during each 10 year service period. If any drywell-to-suppression chamber bypass leak test fails to meet the specified limit, the test schedule for subsequent tests shall be reviewed and approved by the Commission.

Iftwo consecutive tests fail to meet the specified limit, a test shall be performed at least every 18 months until two consecutive tests meet the specified limit, at which time the above test schedule may be resumed.

e.

By conducting a'leakage test on the drywell-to-suppression chamber vacuum breakers at a differential pressure of at least 4.3 psi and verifying that the total leakage area A/(k)" contributed by all vacuum breakers is less than or equal to 30% of the specified limit and the leakage area for an individual set of vacuum breakers is less than or equal to 12% of the specified limit. The vacuum breaker leakage test shall be conducted during each refueling outage for which the:

drywell-to-suppression chamber bypass leak test in Specification 4.6.2.1.d is not conducted.

SUSQUEHANNA - UNIT 2 3/4 6-14 Amendment No. 98

C"NTAI'ADMEN SYS EMS BASES 3/a-6. 2 OEPRESSUR IZATION SYSTEMS Tht sptcfffcatfons of this section tnsurt Chat tht primary contafnmtnt pressure will not txcttd the dtsign pressure of 53 psig during primary systtm b)owoown from ru11 operating pressure.

Tht supprtssion chambtr ~ater providts Cht heat sink for the rtactor coolant system tntrgy reltast following a postulattd rupture of tht system.

The suppres-sion chamber ~ater volume must absorb tht associattd decay and structural stnsiblt htat released during reactor coolant systea blowdown rroa M55 ps)9.

Since all of tht gases in tht 4rywtll are purgt4 into the supprtssion chamber air space during a loss of coolant accfdtnc, tht prtssurt of che lfquf4 must not txcted 53 psfg, the suPPression chamber eaxfee pressure

The design volume of the suppression chawbtr, water an4 air, was obtained by considering that the toCal volt%0 of rtaCtOr COOlant and to be considered is 4ischargtd Co the suppression chawbtr and that tht drywe11 volvo's purged to tht suppression chawber.

Using the afnfmue or maximum water volts given fn thfs specification, contafneent pressurt during tht dtsfgn basis accident fs approxiaattly aS.O psig which is below tht dtsign prtssurt Of 53 psfg.

Naxfaue wattr voiuae of 133,5i0 ft> resulCs in a downcomtr submergence of 12 feet and the ainfa~ volume of 122,410 fts results in a submergtnce approxfeately 21 inches

'less.

The majority of tht 8odtga ttsts wert run with a suboerged length of 4 fetC and with coapltte con4tnsatfon.

Thus, wfth resptct to the downcomer submtrgtnct, Chf S SpeCi fiCatiOn fS adequate.

The eaXfaua temperature at tht end Of Cht blOw-dOwn taeted during tht HuebOldt Say and SOdega Say teStS was 170 F and ChiS iS constrvatively taken to be Cht lfait for coaplttt condensatfon of the reactor

coolant, although condensatfon would occur for teeperatures abovt 170 F.

Should ft be necessary to make the suppression chaaber inoperablt, this shall only be done as sptcffied fn Specification 3.5.3.

Under full power operating condftfons, blowdown f~ an initial suppression chaseer wattr teaperaturt of 90tF rtsults fn a water teeperature of approx-imately 128'F fmedfacely followfng biowdown whfch fs baler the 170'F ustd for coepltte condtnsatfon vfa T-quencher devices.

At this t~eraturt and atmos.

pheriC preSSure, the aVaf lablt NPSH eXCeedS that required by bOth the RHR an4 core spray p~s, thus there fs no dependency on contafreent overpressur t during tht accfdtnt fn)ectfon phase.

If both RNR loops art used for containment cooling.

Chere fs no dependency on contafneent overprtssure for post-LOCA operations.

Experfaental data indicate that excessive steea condensing loads can bt avoided if Che peak local taaporature of the suppression pool is eafntafntd below 200F during any perfo4 of relief valve operation.

Specifications nave been placed on the envtlope of reactor operating con4ftfons so that tht reactor can be deprtssurfzed fn a tfaely canner to avofd the rtgfae of potentially hfgh suppression chaaber loa4fngs.

SUS/UKHANNA - UNIT 2 8 3/a 6 3

CONTAINMENTSYSTEMS BASES DEPRESSURIZATION SYSTEMS Continued)

Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends.

By requiring the suppression pool temperature to be frequently recorded during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken.

The requirement for an external visual examination following any event where potentially high loadings could occur provides assurance that no significant damage was encountered.

Particular attention should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to be the points of highest stress.

In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a safety-relief valve inadvertently opens or sticks open.

As a minimum this action shall include:

(1) use of all available means to close the valve, (2) initiate suppression pool water cooling, (3) initiate reactor shutdown, and (4) if other safety-relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open safety relief valve to assure mixing and uniformity of energy insertion to the pool.

During a LOCA, potential leak paths between the drywell and suppression chamber airspace could result in excessive containment pressures, since the steam flow into the airspace would bypass the heat sink capabilities of the pool.

Potential sources of bypass leakage are the suppression chamber-to-drywall vacuum breakers (VBs), penetrations in the diaphragm floor, and cracks in the diaphragm floor/liner plate and downcomers located in the suppression chamber airspace.

The containment pressure response to the postulated bypass leakage can be mitigated by manually actuating the suppression chamber sprays. An analysis was performed for a design bypass leakage area of A/(k)'~ equal to 0.0535 ft'o verify that the operator has sufficient time to initiate the sprays prior to exceeding the containment design pressure of 53 psig.

The limit of 10% of the design value of 0.0535 ft ensures that the design basis for the steam bypass analysis is met.

The drywell-to-suppression chamber bypass test at a differential pressure of at least 4.3 psi verifies the overall bypass leakage area for simulated LOCA conditions is less than the specified limit. For those outages where the drywall-to-suppression chamber bypass leakage test is not conducted, the VB leakage test verifies that the VB leakage area is less than the bypass limit, with a 70% margin to the bypass limit to accommodate the remaining potential leakage area through the passive structural components.

Previous drywall-to-suppression chamber bypass test data indicates that the bypass leakage through the passive structural components will be much less than the 70% margin.

The VB leakage limit, combined with the negligible passive structural leakage area, ensures that the drywell-to-suppression chamber bypass leakage limit is met for those outages for which the drywell-to-suppression chamber bypass test is not scheduled.

3/4.6.3 PRIMARY CONTAINMENTISOLATIONVALVES The OPERABILITY of the primary containment isolation valves ensures that the containment atmosphere willbe isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment and is consistent with the requirements of GDC 54 through 57 of Appendix A to 10CFR 50.

Containment isolation within the time limits specified for those isolation valves designed to close automatically SUSQUEHANNA - UNIT 2 B 3/4 6-4 Amendment No.$9>

98

CONTAINMENTSYSTEMS BASES DEPRESSURIZATION SYSTEMS Continued)

Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends.

By requiring the suppression pool temperature to be frequently recorded during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken.

The requirement for an external visual examination following any event where potentially high loadings could occur provides assurance that no significant damage was encountered.

Particular attention should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to be the points of highest stress.

In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a safety-relief valve inadvertently opens or sticks'open.

As a minimum this action shall include:

(1) use of all available means to close the valve, (2) initiate suppression pool water cooling, (3) initiate reactor shutdown, and (4) if other safety-relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open safety relief valve to assure mixing and uniformity of energy insertion to the pool ~

During a LOCA, potential leak paths between the drywell and suppression chamber airspace could result in excessive containment pressures, since the steam flow into the airspace would bypass the heat sink capabilities of the pool.

Potential sources of bypass leakage are the suppression chamber-to-drywell vacuum breakers (VBs), penetrations in the diaphragm floor, and cracks in the diaphragm floor/liner plate and downcomers located in the suppression chamber airspace.

The containment pressure response to the postulated bypass leakage can be mitigated by manually actuating the suppression chamber sprays. An analysis was performed for a design bypass leakage area of A/(k)'~equal to 0.0535 ft'o verify that the operator has sufficient time to initiate the sprays prior to exceeding the containment design pressure of 53 psig.

The limit of 10% of the design value of 0.0535 ft'nsures that the design basis for the steam bypass analysis is met.

The drywell-to-suppression chamber bypass test at a differential pressure of at least 4.3 psi verifies the overall bypass leakage area for simulated LOCA conditions is less than the specified limit. For those outages where the drywell-to-suppression chamber bypass leakage test is not conducted, the VB leakage test verifies that the VB leakage area is less than the bypass limit, with a 70% margin to the bypass limit to accommodate the remaining potential leakage area through the passive structural components.

Previous drywell-to-suppression chamber bypass test data indicates that the bypass leakage through the passive structural components will be much less than the 70% margin.

The VB leakage limit, combined with the negligible passive structural leakage area, ensures that the drywell-to-suppression chamber bypass leakage limit is met for those outages for which the drywell-to-suppression chamber bypass test is not scheduled.

P 3/4.6.3 PRIIVIARYCONTAINMENTISOLATIONVALVES The OPERABILITY of the primary containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment and is consistent with the requirements of GDC 54 through 57 of Appendix A to 10CFR 50.

Containment isolation within the time limits specified for those isolation valves designed to close automatically SUSQUEHANNA - UNIT 2 8 3/4 6-4 Amendment No. $9 ~

CONTAINIVIENTSYSTEMS BASES ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.

3 4.6.4 VACUUM RELIEF Vacuum relief breakers are provided to equalize the pressure between the suppression chamber and drywell. This system willmaintain the structural integrity of the primary containment under conditions of large differential pressures.

The vacuum breakers between the suppression chamber and the drywell must not be inoperable in the open position since this would allow bypassing of the suppression pool in case of an accident.

There are five pairs of valves to provide redundancy so that operation may continue for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with no more than one pair of vacuum breakers inoperable in the closed position.

3 4.6.6 SECONDARY CONTAINMENT Secondary containment is designed to minimize any ground level release of radioactive material which may result from an accident.

The Reactor Building provides secondary containment during normal operation when the drywall is sealed and in service.

When the reactor is in COLD SHUTDOWN or REFUELING, the drywell may be open and the Reactor Building then becomes the only containment.

Establishing and maintaining a vacuum in the reactor building with the standby gas treatment system once per 18 months, along with the surveillance of the doors, hatches and dampers, is adequate to ensure that there are no violations of the integrity of the secondary containment.

The OPERABILITYof the standby gas treatment system ensures that sufficient iodine removal capability will be available in the event of a LOCA.

The reduction in containment iodine inventory reduces the resulting site boundary radiation doses associated with containment leakage.

The operation of this system and resultant iodine removal capacity are consistent with the assumptions used in the LOCA analyses.

Cumulative operation of the system with the heaters OPERABLE for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> over a 31 day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters.

3 4.6.6 PRIIVIARYCONTAINIVIENTATMOSPHERE CONTROL The OPERABILITYof the systems required for the detection and control of hydrogen gas ensures that these systems will be available to maintain the hydrogen concentration within the primary containment below its flammable limit during post-LOCA conditions.

The drywell and suppression chamber hydrogen recombiner system is capable of controlling the expected hydrogen generation associated with (1) zirconium-water reactions, (2) radiolytic decomposition of water and (3) corrosion of metals within containment.

Drywell hydrogen mixing is provided to ensure adequate mixing of the containment atmosphere following a LOCA. This mixing action will prevent localized accumulations of hydrogen from exceeding the flammable limit.

The hydrogen control system is consistent with the recommendations of Regulatory Guide 1.7, "Control of Combustible Gas Concentrations in Containment Following a LOCA", March 1971.

SUSQUEHANNA - UNIT 2 B 3/4 6-5 Amendment No.

98