ML20067B578

Amends 176 & 167 to Licenses DPR-77 & DPR-79,respectively, Incorporating Reference to Test Requirements of 10CFR50, App J

ML20067B578
Person / Time
Site:	Sequoyah
Issue date:	02/10/1994
From:	Hebdon F Office of Nuclear Reactor Regulation
To:
Shared Package
ML20067B581	List: ML20067B578 ML20067B589
References
NUDOCS 9402240372
Download: ML20067B578 (39)

Text

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E UNITED STATES 5:

NUCLEAR REGULATORY COMMISSION t

e" WASHINGTON, D.C. 20555-0001 I

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TENNESSEE VALLEY AUTHORITY DOCKET N0.50-32Z SE000YAH NUCLEAR PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 176 License No. DPR-77 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated March 10, 1993, which was amended by letter dated January 31, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and 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 rules and regulations of the Commission; C.

There is reasonable assurance (1) 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; 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.

9402240372 940210 PDR ADOCK 05000327 p

PDR

, 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(?) of Facility Operating License No. DPR-77 is hereby amended to read as follows:

(2) Technical Speciiications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.176, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance, to be implemented within 45 days.

FOR THE NUCLEAR REGULATORY COMMISSION

/ llWW C-Q r

Frederick J. Hebdon, Director Project Directorate II-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

February 10, 1994 P

r

ATTACHMENT TO LICENSE AMENDMENT NO. 176 FACILITY OPERATING LICENSE NO. DPR-77 DOCKET N0. 50-327 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

REMOVE INSERT VII VII 1-2 1-2 3/4 6-1 3/4 6-1 3/4 6-2 3/4 6-2 3/4 6-3 3/4 6-3 3/4 6-4 3/4 6-4 3/4 6-8 3/4 6-8 3/4 6-11 3/4 6-11 3/4 6-12 3/4 6-12 3/4 6-15 3/4 6-15 3/4 6-23 3/4 6-23 B3/4 6-1 83/4 6-1 B3/4 6-2 B3/4 6-2 83/4 6-2a 4

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INDEX 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 3/4.5.2 ECCS SUBSYSTEMS - T,y, greater than or equal to 350*F 3/4 5-4 3/4.5.3 ECCS SUBSYSTEMS - T,y, less than 350*F 3/4 5-8 3/4.5.4 D E L ET ED.........................

3 / 4 5-10 3/4.5.5 REFUELING WATER STORAGE TANK..............

3/4 5-11 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 Containment Air Locks..................

3/4 6-7 Internal Pres sure....................

3/4 6 Ai r 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 Containment Ventilation System.............

3/4 6-15 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray Subsystems 3/4 6-16 Lower Containment Vent Coolers 3/4 6-16b SEQUOYAH - UNIT 1 VII Amendment No. 67, 69, 116, 140, 150, 176

i DEFINITIONS CHANNEL FUNCTIONAL TEST i

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 j

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 l

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 as provided in Table 3.6-2 of Specification 3.6.3.

b.

All equipment hatches are closed and sealed.

c.

Each air lock is in compliance with the requirements of 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 0-rings) is OPERABLE, and f.

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

CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel.

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

CORE OPERATING LIMIT REPORT 1.1) 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 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

1 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT l

l

[0NTAINMENT INTEGRITY i

LIMITING C0_NDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

At least once per 31 days by verifying that all penetrations

• not a.

capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except as provided in Table 3.6-2 of Specification 3.6.3.

b.

By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3.

Perform required visual examinations and leakage rate testing at P, c.

in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions. The maximum allowable leakage rate, L is 0.25% of containment air weight per day at the calculated pe,,ak containment pressure P,,12 psig.

• Except valves, blind flanges, and deactivated automatic valves which are i

located inside the annulus or containment and are locked, sealed or otherwise secured in the closed position.

These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.

SEQUOYAH - UNIT 1 3/4 6-1 Amendment No. 12,130, 176

CONTAINMENT SYSTEMS SECONDARY CONTAINMENT BYPASS LEAKAGE l

LIMITING CONDITION FOR OPERATION 3.6.1.2 Secondary Containment bypass leakage rates shall be limited to a combined bypass leakage rate of less than or equal to 0.25 L, for all penetrations identified in Table 3.6-1 as secondary containment BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING when pressurized to P,.

APPLICABILITY: MODES 1, 2, 3 and 4.

Ell 03:

With the combined bypass leakage rate exceeding 0.25 L for BYPASS LEAKAGE PATHSTOTHEAUXILIARYBUILDING,restorethecombinedbypassleakageratefrom BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING to less than or equal to within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 0.25 L, COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

and in i

SEQUOYAH - UNIT 1 3/4 6-2 Amendment No. 12, 71, 176 l

CONTAINMENT SYSTEMS SECONDARY CONTAINMENT BYPASS LEAKAGE l

SURVElllANCE REQUIREMENTS 4.6.1.2 The secondary containment bypass leakage rates shall be demonstrated:

a.

The combined bypass leakage rate to the auxiliary building shall be and C tests at least once per 24 months excepI. by applicable Type B determined to be less than or equal to 0.25 L for penetrations which are not individually testable; penetrations not individually testable shall be determined to have no detectable leakage when tested with soap bubbles while the containment is pressurized to P, (12 psig) during each Type A test.*

b.

Leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J,Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 P (13.2 psig) and the seal system capacity is adequate to maintain sy, stem pressure (or fluid head for the containment spray system and RHR spray system valves at penetrations 48A, 48B, 49A and 49B) for at least 30 days.

c.

The provisions of Specification 4.0.2 are not applicable,

• Results shall be evaluated against the acceptance criteria of Specifica-tion 4.6.1.1.c in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions.

SEQUOYAH - UNIT 1 3/4 6-3 Amendment No. 12,71,101,102, 127,130, 176

l This page intentionally deleted SEQUOYAH - UNIT 1 3/4 6-4 Amendment No. 12, 71, 101, 130, 176

CONTAINMENT SYSTEMS SVRVElllANCE RE0VIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

After each opening, except when the air lock is being used for a.

multiple entries, then at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, by verifying seal leakage less than or equal to 0.01 L as determined by precision flow measurements when measured for at 'le,ast two minutes with the volume between the door seals at a pressure greater than or equal to 6 psig, b.

By conducting an overall air lock leakage test at not less than P, (12 psig) and by verifying the overall air lock leakage rate is within the limit of Specification 3.6.1.3.b and-the results evaluated in accordance with'10 CFR 50, Appendix J, as modified by approved exemptions:#

v 1.

At least once per six months, and 2.

Prior to establishing CONTAINMENT INTEGRITY if opened when CONTAINMENT INTEGRITY was not required when maintenance has been performed on the air lock that could affect the air lock sealing capability.*

At least once per 6 months by verifying that only one door in each c-.

air lock can be opened at a time.

-l

+

1. The provisions of Specification 4.0.2 are not applicable.

• Exemption to Appendix "J" of 10 CFR 50.

SEQUOYAH - UNIT 1 3/4 6-8 Amendment No. 48,'176 3

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CONTAINMENT SYSTEMS CONTAINMENT VESSEL STRUCTURAL INTEGRITY M TING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containment vessel'shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the structural integrity of the containment vessel not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200*F.

SURVEILLANCE RE0VIREMENTS 4.6.1.6 The structural integrity of the containment vessel shall be determined during the shutdown fir each Type A containment leakage rate test (Specifica-tion 4.6.1.1.c) by a visual inspection of the exposed accessible interior and exterior surfaces of the vessel.

This inspection shall be performed prior to the Type A containment leakage rate test to verify no apparent changes in appearance of the surfaces or other abnormal degradation. Any abnormal degradation of the containment vesse! detected during the above required inspections shall be reported to the Commission pursuant to Specifica-tion 6.6.1.

SEQUOYAH - UNIT 1 3/4 6-11 Amendment No. 36,176 y

CONTAINMENT SYSTEMS SHIELD BUILDING STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.7 The structural integrity of the shield building shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.7.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the structural integrity of the shield building not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200*F.

SURVEILLANCE REQUIREMENTS 4.6.1.7 The structural integrity of the shield building shall be determined during the shutdown for each Type A containment leakage rate test (Specifica-tion 4.6.1.1.c) by a visual inspection of the exposed accessible interior and exterior surfaces of the shield building and verifying no apparent changes in appearance of the concrete surfaces or other abnormal degradation. Any abnormal degradation of the shield building detected during the above required inspections shall be reported to the Commission pursuant to Specifica-tion 6.6.1.

1 i

SEQUOYAH - UNIT 1 3/4 6-12 Amendment No. 36,176

CONTAINMENT SYSTEMS CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.9 One pair (one purge supply line and one purge exhaust line) of containment purge system lines may be open; the containment purge supply and exhaust isolation valves in all other containment purge lines shall be closed.

Operation with purge supply or exhaust isolation valves open for either purging or venting shall be limited to less than or equal to 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> per 365 days.

The 365 day cumulative time period will begin every January 1.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a.

With a purge supply or exhaust isolation valve open in excess of the above cumulative limit, or with more than one pair of containment purge system lines open, close the isolation valve (s) in the purge line(s) within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, b.

With a containment purge supply and/or exhaust isolation valve having a measured leakage rate in excess of 0.05 L,, restore the inoperable valve to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, otherwise be in at least H0T STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6..l.9.1 The position of the containment purge supply and exhaust isolation valves shall be determined at least once per 31 days.

4.6.1.9.2 The cumulative time that the purge supply and exhaust isolation valves are open over a 365 day period shall be determined at least once per 7 days.

4.6.1.9.3 At least once per 3 months, each containment purge supply and exhaust isolation valve shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than or equal to 0.05 L,.*

• Results shall be evaluated against the acceptance criteria of Specifica-tion 4.6.1.1.c in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions.

SEQUOYAH - UNIT 1 3/4 6-15 Amendment No. 18, 120, 176 ww,-evv 7

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TABLE 3.6-2 (Continuedl CONTAINMENT ISOLATION VALVES v,

E8 VALVE NUMBER FUNCTION MAXIMUM ISOLATION TIME (Seconds) 5 C.

PHASE "A" CONTAINMENT VENT ISOLATION (Cont.)

E 13.

FCV-30-50 Upper Compt Purge Air Exh 4*

Z 14.

FCV-30-51 Upper Compt Purge Air Exh 4*

15.

FCV-30-52 Upper Compt Purge Air Exh 4*

16.

FCV-30-53 Upper Compt Purge Air Exh 4*

17.

FCV-30-56 Lower Compt Purge Air Exh 4*

18.

FCV-30-57 Lower Compt Purge Air Exh 4*

19.

FCV-30-58 Inst Room Purge Air Exh 4*

20.

FCV-30-59 Inst Room Purge Air Exh 4*

21.

FCV-90-107 Cntmt Bldg LWR Compt Air Hon 5*

22.

FCV-90-108 Cntmt Bldg LWR Compt Air Hon 5*

23.

FCV-90-109 Cntmt Bldg LWR Compt Air Mon 5*

24.

FCV-90-110 Cntmt Bldg LWR Compt Air Hon 5*

w 25.

FCV-90-111 Cntmt Bldg LWR Compt Air Mon 5*

1 26.

FCV-90-ll3 Cntmt Bldg UPR Compt Air Hon 5*

p 27.

FCV-90-Il4 Cntmt Bldg UPR Compt Air Mon 5*

ro 28.

FCV-90-ll5 Cntmt Bldg UPR Compt Air Hon 5*

29.

FCV-90-Il6 Cntmt Bldg UPR Compt Air Mon 5*

30.

FCV-90-117 Cntmt Bldg UPR Compt-Air Mon 5*

D.

OTHER

@R 1.

FCV-30-46 Vacuum Relief Isolation Valve 25 2

2.

FCV-30-47 Vacuum Relief Isolation Valve 25 5

3.

FCV-30-48 Vacuum Relief Isolation Valve 25 4.

FCV-62-90 Normal Charging Isolation Valve 12 2

• Provisions of LCO 3.0.4 are not applicable if valve is secured in its isolated position with power removed Ei and leakage limits of Specification 4.6.1.1.c are satisfied.

For purge valves, leakage limits under Surveillance Requirement 4.6.1.9.3 must also be satisfied.

R!

~

1. Provisions of LCO 3.0.4 are not applicable if valve is secured in its isolated position with power removed g

and either FCV-62-73 or FCV-62-74 is maintained operable.

• • This valve is required after completion of the associated modification.

3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT The safety design basis for primary contait. ment is that the containment must withstand the pressures and temperatures of the limiting design basis accident (DBA) without exceeding the design leakage rates.

The DBAs that result in a challenge to containment OPERABILITY from high pressures and temperatures are a loss of coolant accident (LOCA), a steam line break, and a rod ejection accident (REA).

In addition, release of significant fission product radioactivity within containment can occur from a LOCA or REA.

In the DBA analyses, it is assumed that the containment is OPERABLE such that, 1

for the DBAs involving release of fission product radioactivity, release to the environment is controlled by the rate of containment leakage.

This leakage rate limitation will limit the site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions.

The containment was designed with an allowable leakage rate of 0.25 percent of containment air weight per day.

This leakage rate, used in the evaluation of offsite doses resulting from

{

accidents, is defined in 10 CFR 50, Appendix J, as L : the maximum allowable containment leakage rate at the calculated peak containment internal pressure (P

resulting from the limiting DBA.

by,)L, forms the basis for the acceptance criteria imposed on all containmentThe al leakage rate testing.

L i analysis at P - 12.0 psfg.s assumed to be 0.25 percent per day in the safety As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L during performance of the periodic tests to account for possible degradati,n of o

the containment leakage barriers between tests.

Primary containment INTEGRITY or operability is maintained by limiting leakage to within the acceptance criteria of 10 CFR 50, Appendix J.

Individual leakage rates specified for the containment air lock (LCO 3.6.1.3), purge valves (LCO 3.6.1.9) and secondary bypass leakage (LCO 3.6.1.2) are not specifically part of the acceptance criteria of 10 CFR 50, Appendix J.

Therefore, leakage rates exceeding these individual limits do not result in the primary containment being inoperable unless the leakage, when combined with other Type B and C test leakages, exceeds the acceptance criteria of Appendix J.

3/4.6.1.2 SECONDARY CONTAINMENT BYPASS LEAKAGE The safety design basis for containment leakage assumes that 75 percent of the leakage from the primary containment enters the shield building annulus for filtration by the emergency gas treatment system. The remaining 25 percent of the primary containment leakage, which is considered to be bypassed to the auxiliary building, is assumed to exhaust directly to the atmosphere without filtration during the first 5 minutes of the accident. After 5 minutes, any bypass leakage to the auxiliary building is filtered by the auxiliary buildtng gas treatment system. A tabulation of potential secondary containment bypass SEQUOYAH - UNIT 1 B 3/4 6-1 Amendment No. 102, 127, 176

3/4.6 CONTAINMENT SYSTEMS BASES leakage paths to the auxiliary building is provided in Table 3.6-1. Restricting the leakage through the bypass leakage paths in Table 3.6-1 to 0.25 L provides assurance that the leakage fraction assumptions used in the evaluatio,n of site boundary radiation doses remain valid.

3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate. Surveillance testing of the air lock seals provide assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.

3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that 1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the annulus atmosphere of 0.5 psig and 2) the containment peak pressure does not exceed the maximum allowable internal pressure of 12 psig during LOCA conditions.

3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that 1) the containment air mass is limited to an initial mass sufficiently low to prevent exceeding the maximum allowable internal pressure during LOCA conditions and

2) the ambient air temperature does not exceed that temperature allowable for the continuous duty rating specified for equipment and instrumentation located within containment.

The containment pressure transient is sensitive to the initially contained air mass during a LOCA.

The contained air mass increases with decreasing temperature.

The lower temperature limits of 100*F for the lower compartment, 85'F for the upper compartment, and 60*F when less than or equal to 5% of RATED THERMAL POWER will limit the peak pressure to an acceptable value. The upper temperature limit influences the peak accident temperature slightly during a LOCA; however, this limit is based primarily upon equipment protection and anticipated operating conditions. Both the upper and lower temperature limits are consistent with the parameters used in the accident analyses.

3/4.6.1.6 CONTAINMENT VESSEL STRUCTVRAL INTEGRITY This limitation ensures that the structural integrity of the containment steel vessel will be maintained comparable to the original design standards for the life of the facility.

Structural integrity is required to ensure that the vessel will withstand the maximum pressure of 12 psig in the event of a LOCA. A visual inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability.

SEQUOYAH - UNIT 1 B 3/4 6-2 Amendment No. 102, 127, 176

I 3/4.6 CONTAINMENT SYSTEMS BASES 3/4,6,1.7 SHIELO BUILDING STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment shield building will be maintained comparable to the original design standards for the life of the facility.

Structural integrity is required to provide 1) protection for the steel vessel from external missiles, 2) radiation shielding in the event of a LOCA, and 3) and annulus surrounding the steel vessel that can be maintained at a negative pressure during accident conditions.

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UNITED STATES iik ~

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NUCLEAR REGULATORY COMMISSION

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WASHINGTON, D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327 i

i SEOU0YAH NUCLEAR PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE I

Amendment No. 167 License No. DPR-77 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated March 10, 1993, which was amended by letter dated January 31, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and 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 rules and 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; 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.

t

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of facility Operating License No. DPR-79 is hereby amended to read as follows:

(2) Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 167, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance, to be implemented within 45 days.

FOR THE NUCLEAR REGULATORY COMMISSION L hlG& D D

Frederick J. Hebdon, Director Project Directorate II-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation y

Attachment:

Changes to the Technical Specifications i

Date of Issuance:

February 10, 1994

-l

ATTACHMENT TO LICENSE AMENDMENT NO 167 FACILITY OPERATING LICENSE NO. OPR-79 DOCKET NO. 50-328 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.

The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

REMOVE INSERT VII VII l-2 1-2 3/4 6-1 3/4 6-1 3/4 6-2 3/4 6-2 3/4 6-3 3/4 6-3 3/4 6-4 3/4 6-4 3/4 6-8 3/4 6-8 3/4 6-11 3/4 6-11 3/4 6-12 3/4 6-12

-l 3/4 6-15 3/4 6-15 3/4 6-23 3/4 6-23 B3/4 6-1 B3/4 6-1 B3/4 6-2 B3/4 6-2 B3/4 6-2a i

)

1 4

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTIQB PAGE 3/L5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS j

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

3/4 5-1

]

Deleted.........................

3/4 5-3 3/4.5.2 ECCS SUBSYSTEMS - T,y, greater than or equal to 350*F 3/4 5-4 3/4.5.3 ECCS SUBSYSTEMS - T,y, less than 350*F 3/4 5-8 3/4.5.4 DELETED.......................... 3/4 5-10 1

3/4.5.5 REFUELING WATER STORAGE TANK..............

3/4 5-11 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 l

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

3/4 6-7 Internal Pres sure..................... 3/4 6-9 i

Air Temperature...................... 3/4 6-10 3

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 Containment Ventilation System 3/4 6-15 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray Subsystems 3/4 6-16 Lower Containment Vent Coolers 3/4 6-16b SEQUOYAH - UNIT 2 VII Amendment No. 59, 61, 131, 140, 167

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 1

alarm and/or trip functions.

b.

Bistable channels - the injection of a simulated signal into the i

sensor to verify OPERABILITY including alarm and/or trip functions.

l c.

Digital channels - the injection of a simulated signal into the chan-nel as close to the sensor input to the process racks as practicable to verify 0PERABILITY including alarm and/or trip functions.

CONTAINMENT INTEGRITY i

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 as provided in Table 3.6-2 of Specification 3.6.3.

b.

All equipment hatches are closed and sealed.

l c.

Each air lock is in compliance with the requirements of Specification

)

3.6.1.3,

]

d.

The containment leakage rates are within the limits of Specifica-tion 4.6.1.1.c, i

e.

The sealing mechansim associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE, and f.

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

1 CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATLQH 1.9 CORE ALTERATION shall be the movement or manipulation of any component within'the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative 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.

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 2 1-2 Amendment No. 63,117,132,146, 167

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY:

MODES 1, 2, 3 and 4.

ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT. INTEGRITY within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a.

At least once per 31 days by verifying that all penetrations

• not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except as provided in Table 3.6-2 of Specification 3.6.3.

b.

By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3.

Perform required visual examinations and leakage rate testing at P, c.

in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions. The maximum allowable leakage rate, L, is 0.25% of containment air weight per day at the calculated p,eak containment pressure P,12 psig.

• Except valves, blind flanges, and deactivated automatic valves which are located inside the annulus or containment and are locked, sealed or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.

SEQUOYAH - UNIT 2 3/4 6-1 Amendment No. 117, 167

~.

_ ~

CONTAINMENT SYSTEMS SECONDARY CONTAINMENT BYPASS LEAKAGE l

LIMITING CONDITION FOR OPERATION 3.6.1.2 Secondary Containment bypass leakage rates shall be limited to a combined bypass leakage rate of less than or equal to 0.25 L, for all penetrations identified in Table 3.6-1 as secondary containment BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING when pressurized to P,.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the combined bypass leakage rate exceeding 0.25 L for BYPASS LEAKAGE PATHSTOTHEAUXILIARYBUILDING,restorethecombinedbypassleakageratefrom BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING to less than or equal to 0.25 L within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least H0T STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in C6LD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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i SEQUOYAH - UNIT 2 3/4 6-2 Amendment No. 63, 167

CONTAINMENT SYSTEMS S1(QNDARY CONTAINMENT BYPASS LEAKAGE l

SURVEllLANCE REQUIREMENTS 4.6.1.2 The secondary containment bypass leakage rates shall be demonstrated:

I a.

The combined bypass leakage rate to the auxiliary building shall be determined to be less than or equal to 0.25 L by applicable Type B andCtestsatleastonceper24monthsexceptforpenetrationswhich are not individually testable; penetrations not indivioually testable shall be determined to have no detectable leakage when tested with soap bubbles while the containment is pressurized to P,, (12 psig) during each Type A test.*

b.

Leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J,Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 P,in (13.2 psig) and the seal system capacity is adequate to mainta system pressure (or fluid head for the containment spray system and RHR spray system valves at penetrations 48A, 48B, 49A and 49B) for at least 30 days.

c.

The provisions of Specification 4.0.2 are not applicable.

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• Results shall be evaluated against the acceptance criteria of Specifica-tion 4.6.1.1.c in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions.

SEQUOYAH - UNIT 2 3/4 6-3 Amendment No. 63,90,104,117,126, 167

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8 SEQUOYAH - UNIT 2 3/4 6-4 Amendment No. 63,90,104,117,126, 167

CONTAINMENT SYSTEMS SURVEILLANCE RE0VIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

a.

After each opening, except when the air lock is being used for multiple entries, then at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, by verifying seal leakage less than or equal to 0.01 L as determined by precision flow i

measurements when measured for at le,ast two minutes with the volume between the door seals at a pressure greater than or equal to 6 psig, b.

By conducting an overall air lock leakage test at not less than P, (12 psig) and by verifying the overall air lock leakage rate is within the limit of Specification 3.6.1.3.b and the results evaluated in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions:#

1.

At least once per six months, and 2.

Prior to establishing CONTAINMENT INTEGRITY if opened when CONTAINMENT INTEGRITY was not required when maintenance has been performed on the air lock that could affect the air lock sealing capability.*

c.

At least once per 6 months by verifying that only one door in each air lock can be opened at a time.

1. The provisions of Specification 4.0.2 are not applicable.

• Exemption to Appendix "J" of 10 CFR 50.

SEQUOYAH - UNIT 2 3/4 6-8 Amendment No. 40,167

CONTAINMENT SYSTEMS CONTAINMENT VESSEL STRUCTURAL INTEGRITY llMITING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containment vessel shall be maintained at'a level consistent with the acceptance criteria in Specification 4.6.1.6.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the structural integrity of the containment vessel not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200*F.

SURVEILLANCE REQUIREMENTS 4.6.1.6 The structural integrity of the containment vessel shall be determined during the shutdown for each Type A containment leakage rate test (Specifica-tion 4.6.1.1.c) by a visual inspection of the exposed accessible interior and exterior surfaces of the vessel. This inspection shall be performed prior to the Type A containment leakage rate test to verify no apparent changes in appearance of the surfaces or other abnormal degradation. Any abnormal degradation of the containment vessel detected during the above required inspections shall be reported to the Commission pursuant to Specifica-tion 6.6.1.

}

4 SEQUOYAH - UNIT 2 3/4 6-11 Amendment-No. 2& 167

CONTAINMENT SYSTEMS SHIELD BUILDING STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.7 The structural integrity of the shield building shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.7.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the structural integrity of the shield building not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200'F.

SURVEILLANCE REQUIREMENTS 4.6.1.7 The structural integrity of the shield building shall be determined during the shutdown for each Type A containment leakage rate test (Specifica-tion 4.6.1.1.c) by a visual inspection of the exposed accessible interior and exterior surfaces of the shield building and verifying no apparent changes in appearance of the concrete surfaces or other abnormal degradation. Any abnormal degradation of the shield building detected during the above required inspections shall be reported to the Commission pursuant to.Specifica-tion 6.6.1.

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1 SEQUOYAH - UNIT 2 3/4 6-12 Amendment No. 28,167

=. _

CONTAINMENT SYSTEMS CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.9 One pair (one purge supply line and one purge exhaust line) of containment purge system lines may be open; the containment purge supply and exhaust isolation valves in all other containment purge lines shall be closed.

Operation with purge supply or exhaust isolation valves open for either purging or venting shall be limited to less than or equal to 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> per 365 days.

The 365 day cumulative time period will begin every January 1.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACT10ti:

a.

With a purge supply or exhaust isolation valve open in excess of the above cumulative limit, or with more than one pair of containment purge system lines open, close the isolation valve (s) in the purge line(s) within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.

With a containment purge supply and/or exhaust isolation valve having a measured leakage rate in excess of 0.05 L, restore the inoperable valve to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.1.9.1 The position of the containment purge supply and exhaust isolation valves shall be determined at least once per 31 days.

4.6.1.9.2 The cumulative time that the purge supply and exhaust isolation valves are open over a 365 day period shall be determined at least once per 7 days.

4.6.1.9.3 At least once per 3 months, each containment purge supply and exhaust isolation valve shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than or equal to 0.05 L,.*

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• Results shall be evaluated against the acceptance criteria of Specifica-tion 4.6.1.1.c in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions.

SEQUOYAH - UNIT 2 3/4 6-15 Amendment No. 9, 109, 167 l

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7

TABLE 3.6-2 (Continued _1 CONTAINMENT ISOLATION VALVES VALVE NUMBER FUNCTION MAXIMUM ISOLATION TIME (Seconds) i C.

PHASE "A" CONTAINMENT VENT ISOLATION (Cont.)

C5 13.

FCV-30-50 Upper Compt Purge Air Exh 4*

f 14.

FCV-30-51 Upper Compt Purge Air Exh 4*

15. FCV-30-52 Upper Compt Purge Air Exh 4*

16.

FCV-30-53 Upper Compt Purge Air Exh 4*

17. FCV-30-56 Lower Compt Purge Air Exh 4*

18. FCV-30-57 Lower Compt Purge Air Exh 4*

19.

FCV-30-58 Inst Room Purge Air Exh 4*

20.

FCV-30-59 Inst Room Purge Air Exh 4*

21. FCV-90-107 Cntmt Bldg LWR Compt Air Mon 5*

22. FCV-90-108 Cntmt Bldg LWR Compt Air Mon 5*

23.

FCV-90-109 Cntmt Bldg LWR Compt Air Mon 5*

2'

24. FCV-90-110 Cntmt Bldg LWR Compt Air Mon 5*

25. FCV-90-111 Cntmt Bldg LWR Compt Air Mon 5*

i'

26. FCV-90-ll3 Cntmt Bldg UPR Compt Air Mon 5*

O

27. FCV-90-ll4 Cntmt Bldg UPR Compt Air Mon 5*

28.

FCV-90-ll5 Cntmt Bldg UPR Compt Air Mon 5*

29. FCV-90-116 Cntmt Bldg UPR Compt Air Mon 5*

30. FCV-90-117 Cntmt Bldg UPR Compt Air Mon 5*

D.

OTHER k

1.

FCV-30-46 Vacuum Relief Isolation Valve 2

R 2.

FCV-30-47 Vacuum Relief Isolation Valve 2

2 3.

FCV-30-48 Vacuum Ralief Isolation Valve 2

?+

4.

FCV-62-90 Norma'. Charging Isolation Valve 1

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~*

• Provisions of LC0 3.0.4 are not applicable if valve is secured in its isolated position with power removed C

and leakage limits of Specification 4.6.1.1.c are satisfied. For purge valves, leakage limits under l

'y surveillance Requirement 4.6.1.9.3 must also be satisfied.

N

1. Provisions of LCO 3.0.4 are not applicable if valve is secured in its isolated position with power removed and either FCV-62-73 or FCV-62-74 is maintained operable.

• • This valve is required after completion of the associated modification.

3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT The safety design basis for primary containment is that the containment must withstand the pressures and temperatures of the limiting design basis accident (DBA) without exceeding the design leakage rates.

The DBAs that result in a challenge to containment OPERABILITY from high pressures and temperatures are a loss of coolant accident (LOCA), a steam line break, and a rod ejection accident (REA).

In addition, release of significant fission product radioactivity within containment can occur from a LOCA or REA.

In the DBA analyses, it is assumed that the containment is OPERABLE such that, for the DBAs involving release of fission product radioactivity, release to the environment is controlled by the rate of containment leakage. This leakage rate limitation will limit the site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions. The containment was designed with an allowable leakage rate of 0.25 percent of containment air weight per day.

This leakage rate, used in the evaluation of offsite doses resulting from accidents, is defined in 10 CFR 50, Appendix J, as L : the maximum allowable containment leakage rate at the calculated peak centainment internal pressure (P

resulting from the limiting DBA.

by,)L, forms the basis for the acceptance criteria imposed on all containmentThe a leakage rate testing.

L i

-12.0psfg.sassumedtobe0.25percentperdayinthesafety analysis at P As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L during performance of the periodic tests to account for possible degradati,n of o

the containment leakage barriers between tests.

Primary containment INTEGRITY or operability is maintained by limiting leakage to within the acceptance criteria of 10 CFR 50, Appendix J.

Individual leakage rates specified for the containment air lock (LCO 3.6.1.3), purge valves (LCO 3.6.1.9) and secondary bypass leakage (LCO 3.6.1.2) are not specifically part of the acceptance criteria of 10 CFR 50, Appendix J.

Therefore, leakage rates exceeding these individual limits do not result in the primary containment being inoperable unless the leakage, when combined with other Type B and C test leakages, exceeds the acceptance criteria of Appendix J.

3/4.6.1.2 SECONDARY CONTAINMENT BYPASS LEAKAGE The safety design basis for containment leakage assumes that 75 percent of the leakage from the primary containment enters the shield building annulus for filtration by the emergency gas treatment system. The remaining 25 percent of the primary containment leakage, which is considered to be bypassed to the auxiliary building, is assumed to exhaust directly to the atmosphere without filtration during the first 5 minutes of the accident. After 5 minutes, any bypass leakage to the auxiliary building is filtered by the auxiliary building gas treatment system. A tabulation of potential secondary containment bypass SEQUOYAH - UNIT 2 B 3/4 6-1 Amendment No. 91, 139, 167

3/4.6 CONTAINMENT SYSTEMS BASES leakage paths to the auxiliary building is provided in Table 3.6-1. Restricting the leakage through the bypass leakage paths in Table 3.6-1 to 0.25 L provides assurance that the leakage fraction assumptions used in the evaluatio,n of site boundary radiation doses remain valid.

3/4.6.1.3 CONTAINMENT AIR LRCM The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate.

Surveillance testing of the air lock seals provide assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.

3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that 1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the annulus atmosphere of 0.5 psig and 2) the containment peak pressure does not exceed the maximum allowable internal pressure of 12 psig during LOCA conditions.

3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that 1) the containment air mass is limited to an initial mass sufficiently low to prevent exceeding the maximum allowable internal pressure during LOCA conditions and i

2) the ambient air temperature does not exceed that temperature allowable for the continuous duty rating specified for equirment and instrumentation located within containment.

The containment pressure transient is sensitive to the initially contained air mass during a LOCA.

The contained air mass increases with decreasing temperature.

The lower temperature limits of 100*F for the lower compartment, 85*F for the upper compartment, and 60*F when less than or equal to 5% of RATED THERMAL POWER will limit the peak pressure to an acceptable value. The upper temperature limit influences the peak accident temperature slightly during a LOCA; however, this limit is based primarily upon equipment protection and anticipated operating conditions. Both the upper and lower temperature limits are consistent with the parameters used in the accident analyses.

3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment steel vessel will be maintained comparable to the original design standards for the life of the facility.

Structural integrity is required to ensure that the vessel will withstand the maximum pressure of 12 psig in the event of a LOCA. A visual inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability.

SEQUOYAH - UNIT 2 B 3/4 6-2 Amendment No. 91, 139, 167

3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1.7 SHIELD BUILDING STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment shield building will be maintained comparable to the original design standards for the life of the facility.

Structural integrity is required to provide 1) protection for the steel vessel from external missiles, 2) radiation shielding in the event of a LOCA, and 3) and annulus surrounding the steel vessel that can be maintained at a negative pressure during accident conditions.

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SEQUOYAH - UNIT 2 B 3/4 6-2a Amendment No. 91, 139, 167-

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