ML20085G574
| ML20085G574 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 06/13/1995 |
| From: | Hebdon F NRC (Affiliation Not Assigned) |
| To: | TENNESSEE VALLEY AUTHORITY |
| Shared Package | |
| ML20085G578 | List: |
| References | |
| NUDOCS 9506200274 | |
| Download: ML20085G574 (35) | |
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4 UNITED STATES s
g NUCLEAR REGULATORY COMMISSION t
WASHINGTON, D.C. 20555-0001 49.,....,o TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327 SE000YAH NUCLEAR PLANT. UNIT 1
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AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 203 License No. DPR-77 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated April 6,1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Comission'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 Comission; 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 Comission'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 Comission's regulations and all applicable requirements have been satisfied.
9506200274 950613 PDR ADOCK 050003 7
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. 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-77 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 203, 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 60 days.
FOR THE NUCLEAR REGULATORY COMMISSION ne
-- $c Frederick J. Hebdbn, Director Project Directorate II-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
June 13, 1995 i
4
Mr. Oliver D. Kingsley, Jr.
SEQUOYAH NUCLEAR PLANT 4
Tennessee Valley Authority cc:
Mr. O. J. Zeringue, Sr. Vice President TVA Representative Nuclear Operations Tennessee Valley Authority Tennessee Valley Authority 11921 Rockville Pike 3B Lookout Place Suite 402 1101 Market Street Rockville, MD 20852 Chattanooga, TN 37402-2801 Regional Administrator Dr. Mark 0. Medford, Vice President U.S. Nuclear Regulatory Commission Engineering & Technical Services Region II Tennessee Valley Authority 101 Marietta Street, NW., Suite 2900 3B Lookout Place Atlanta, GA 30323 1101 Market Street Chattanooga, TN 37402-2801 Mr. William E. Holland Senior Resident Inspector Mr. D. E. Nunn, Vice President Sequoyah Nuclear Plant New Plant Completion U.S. Nuclear Regulatory Commission Tennessee Valley Authority 2600 Igou Ferry Road 3B Lookout Place Soddy Daisy, TN 37379 1101 Market Street Chattanooga, TN 37402-2801 Mr. Michael H. Mobley, Director Division of Radiological Health Site Vice President 3rd Floor, L and C Annex Sequoyah Nuclear Plant 401 Church Street Tennessee Valley Authority Nashville, TN 37243-1532 P.O. Box 2000 Soddy Daisy, TN 37379 County Judge Hamilton County Courthouse General Counsel Chattanooga, TN 37402-2801 Tennessee Valley Authority ET llH 400 West Summit Hill Drive Knoxville, TN 37902 Mr. P. P. Carier, Manager Corporate Licensing Tennessee Valley Authority 4G Blue Ridge 1101 Market Street Chattanooga, TN 37402-2801 Mr. Ralph H. Shell Site Licensing Manager Sequoyah Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Soddy Daisy, TN 37379
ATTACHMENT TO LICENSE AMENDMENT NO. 203 FACILITY OPERATING LICENSE NO. DPR-77 DOCKET NO. 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 l-2 1-2 3/4 6-1 3/4 6-1 3/4 6-2 3/4 6-2 3/4 6-4 3/4 6-4 3/4 6-5 3/4 6-6 3/4 6-6a 3/4 6-17 3/4 6-17 3/4 6-18 3/4 6-18 3/4 6-19 3/4 6-19 3/4 6-20 3/4 6-21 3/4 6-22 3/4 6-23 3/4 8-15 3/4 8-15 3/4 8-16 3/4 8-16 3/4 8-17 3/4 8-17 3/4 8-18 3/4 8-18 3/4 8-19 B3/4 6-2 B3/4 6-2 B3/4 6-3 B3/4 6-3 B3/4 6-3a m
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DEFINITIONS CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be:
a.
Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions.
b.
Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.
c.
Digital channels - the injection of a simulated signal into the channel as close to the sensor input to the process racks as practicable to verify OPERABILITY including alarm and/or trip functions.
CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:
a.
All penetrations required to be closed during accident conditions are either:
1)
Capable of being closed by an OPERABLE containment automatic isolation valve system, or 2)
Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are open under administrative control as permitted by Specifica-tion 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.
I CORE ALTERATION 1.9 CORE ALTERATIONS shall be the movement of any fuel, sources, reactivity cortrol components, or other components affecting reactivity within the reactor vessel with the head removed and fuel in the vessel.
Suspension of CORE i
ALTERATIONS shall not preclude completion of movement of a component to a safe position.
CORE OPERATING LIMIT REPORT i
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 1 1-2 Amendment No. 12,71,130,141,155, 176, 201, 203
o 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 REOUIREMENTS 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 for valves that are open under administrative control as permitted by Specification 3.6.3.
b.
By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3.
c.
Perform required visual examinations and leakage rate testing at P, in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions. The maximum allowable ~ 1eakage rate, L is 0.25% of containment air weight per day at the calculated p, k containment ea pressure P,,12 psig.
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- Except valves, blind flanges, and deactivated automatic valves which are located inside the annulus or containment or the main steam valve vaults 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, 191, 203
CONTAINMENT SYSTEMS SECONDARY CONTAINMENT BYPASS LEAKAGE LINITING 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 penetrationsthataresecondarycontainmentBYPASSLEAKAGEPATHSTOTHE l
AUXILIARY BUILDING when pressurized to P,.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With the combined bypass leakage rate exce'eding 0.25 L for BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING, restore the combined bypass leakage rate from 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 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 />.
SEQUOYAH - UNIT 1 3/4 6-2 Amendment No. 12, 71, 176, 203
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Pages 3/4 6-4 through 3/4 6-6a intentionally deleted l
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SEQUOYAH - UNIT 1 3/4 6-4 Amendment No.12, 71'3 101, 130, 176, 20
1 CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 Each containment isolation valve shall be OPERABLE.*
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' APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
a.
With one or more of the isolation valve (s), except containment vacuum relief isolation valve (s), inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and either:
1.
Restore the inoperable valve (s) to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or 2.
Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one deactivated automatic valve secured in the isolation position, or 3.
Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual valve or blind flange; or 4.
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 SHVTDOWN 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 one or more containment vacuum relief isolation valve (s) inoperable, I
the valve (s) must be returned to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 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 /> 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 />.
c.
The provisions of Specification 3.0.4 do not apply.
g SVRVElllANCE RE0VIREMENTS 4.6.3.1 Deleted l
- Penetration flow path (s) may be unisolated intermittently under administrative controls.
SEQUOYAH - UNIT 1 3/4 6-17 Amendment No. 12, 203
CONTAINMENT SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued) 4.6.3.2 Each automatic containment isolation valve shall be demonstrated l
OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by:
a.
Verifying that on a Phase A containment isolation test signal, each Phase A isolation valve actuates to its isolation position.
b.
Verifying that on a Phase B containment isolation test signal,' each Phase B isolation valve actuates to its isolation position.
c.
Verifying that on a Containment Ventilation isolation test signal, each Containment Ventilation Isolation valve actuates to its isolation position.
d.
Verifying that on a high containment pressure isolation test signal, each Containment Vacuum Relief Valve actuates to its isolation position.
l e.
Verifying that on a Safety Injection test signal that the Normal l
Charging Isolation valve actuates to its isolation position.
i 4.6.3.3 The isolation time of each power operated or automatic containment isolation valve shall be determined to be within its limit when tested pursuant to Specification 4.0.5.
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SEQUOYAH - UNIT 1 3/4 6-18 Amendment No. 12, 81, 101, 120 203
i lr Pages 3/4 6-19 through 3/4 6-23 intentionally deleted 1
SEQUOYAH - UNIT 1 3/4 6-19 Amendment No. 37, 70, 145, ISS 203 i
ELECTRICAL POWER SYSTEMS 3/4.8.3 ELECTRICAL E0VIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES LIMITING CONDITION FOR OPERATION 3.8.3.1 Primary and Backup containment penetration conductor overcurrent protective devices associated with each containment electrical penetration circuit shall be OPERABLE. The scope of these protective devices excludes those circuits for which credible fault currents would not exceed the electrical penetration design rating.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With one or more of the containment penetration conductor overcurrent protective devices inoperable:
I a.
Restore the protective device (s) to OPERABLE status or de-energize the circuit (s) by tripping the associated backup circuit breaker within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and verify the backup circuit breaker to be tripped at least once per 7 days thereafter; the provisions of Specification 3.0.4 are not applicable to overcurrent devices in circuits which have their backup circuit breakers tripped, or b.
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 RE0VIREMENTS 4.8.3.1 All containment penetration conductor overcurrent protective devices shall be demonstrated OPERABLE:
a.
At least once per 18 months:
1.
For at least one 6.9 kV reactor coolant pump circuit, such that all reactor coolant pump circuits are demonstrated OPERABLE at least once per 72 months, by performance of:
(a) A CHANNEL CALIBRATION of the associated protective relays specified in appropriate plant instructions, and (b) An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits function as designed.
j SEQUOYAH - UNIT 1 3/4 8-15 Amendment No. 42, 110, 203 j
ELECTRICAL POWER SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued)
(c) For each circuit breaker found inoperable during these functional tests, an additional representative sample of at least 1 of the circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that, type have been functionally tested.
2.
By selecting and functionally testing a representative sample of at least 10% of each type of lower voltage circuit breakers. Circuit breakers selected for functional testing shall be selected on a rotating basis. The functional test shall consist of injecting a current input i
at the specified setpoint to each selected circuit breaker and verifying that each circuit breaker functions as designed. Circuit breakers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation.
For each circuit breaker found inoperable during these functional tests an additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested.
3.
By selecting and verifying a representative sample of each type of fuse on a rotating basis. Verification will be accomplished as described by SR 4.8.3.1.a.3.a.
Each representative sample of fuses shall include at least 10% of all fuses of that type.
Fuses found inoperable during l
verification shall be replaced with OPERABLE fuses prior to resuming operation.
For each fuse found inoperable during verification, an additional representative sample of at least 10% of all fuses of that type shall be functionally tested until no more failures are found or i
all fuses of that type have been functionally tested, i
(a) A fuse verification and maintenance program will be maintained to ensure that:
1.
The proper size and type of fuse is installed, l
2.
The fuse shows no sign of deterioration, and 3.
The fuse connections are tight and clean.
b.
At least once per 60 months by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedures prepared in conjunction with manufacturer's recommendations.
SEQUOYAH - UNIT 1 3/4 8-16 Amendment No. 42, 110, 203
c ELECTRICAL POWER SYSTEMS i
MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION LIMITING CONDITION FOR OPERATION 3.8.3.2 The thermal overload protection devices, integral with the motor starter, of each valve used in safety systems shall be OPERABLE.
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APPLICABILITY: Whenever the motor operated valve is required to be OPERABLE.
ACTION:
With one or more of the thermal overload protection devices inoperable, declare the affected valve (s) inoperable and apply the ACTION Statement to the affected valve (s).
SURVEILLANCE REQUIREMENTS 4.8.3.2 The above required thermal overload protection devices shall be demonstrated OPERABLE:
a.
At least once per 18 months by the performance of a CHANNEL CALIBRATION of a representative sample of at least 25% of all thermal overload devices i
which are not bypassed, such that each non-bypassed device is calibrated at i
least once per 6 years, b.
At least once per 18 months, by the performance of a CHANNEL FUNCTIONAL TEST of the bypass circuitry for those thermal overload devices which are normally in force during plant operation and bypassed under accident conditions.
s SEQUOYAH - UNIT 1 3/4 8-17 Amendment No. 33, 80, 128, 203
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Pages 3/4 8-18 through 3/4 8-19 intentionally deleted E
E SEQUOYAH - UNIT 1 3/4 8-18 Amendment No. 203
l 3/4.6 CONTAINMENT SYSTEMS BASES leakage paths to the auxiliary building is provided in plant procedures.
Restricting the leakage through the bypass leakage paths to 0.25 L provides assurancethattheleakagefractionassumptionsusedintheevaluationofsite 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.
1 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 ex: ceding 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 tha upper and lower tecperature 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 1 B 3/4 6-2 Amendment No. 102, 127, 176, 203
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CONTAINMENT SYSTEMS BASES 3/4.6.1.8 EMERGENCY GAS TREATMENT SYSTEM (EGTS)
The.0PERA8ILITY of the EGTS cleanup subsystem ensures that during LOCA conditions, containment vessel leakage into the annulus will be filtered through the HEPA filters and charcoal adsorber trains prior to discharge to the atmosphere. This requirement is necessary to meet the assumptions used in the accident analyses and limit the site boundary radiation doses to within the limits of 10 CFR 1,00 during LOCA conditions. Cumulative operation of the sys-ten with the heaters on 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 absorbers and HEPA filters. ANSI N510-1975 will be used as a procedural guide for surveillance testing.
3/4.6.1.9 CONTAINMENT VENTILATION SYSTEM use of the containment purge lines is restricted to only one pair (one supply line and one exhaust line) of purge system lines at a time to ensure that the site boundary dose guidelines of 10 CFR Part 100 would not be exceeded in the event of a loss of coolant accident during purging operations. The analysis of this accident assumed purging thraugh the largest pair of lines (a 24 inch inlet line and a 24 inch outlet line), a pre-existing iodine spike in the reactor coolant and four second valve closure times.
3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SUBSYSTEMS The OPERABILITY of the containment spray subsystems ensures that contain-ment depressurization and cooling capability will be available in the event of a LOCA. The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the accident analyses.
3/4.6.2.2 CONTAINMENT COOLING FANS The OPERABILITY of the lower containment vent coolers ensures that ade-quate heat removal capacity is available to provide long-term cooling following a non-LOCA event.
Postaccident use of these coolers ensures containment tem-peratures remain within environmental qualification limits for all safety-related equipment required to remain functional.
3/4.6.3 CONTAINMENT ISOLATION VALVES The operability of containment isolation valves ensures that the l
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.
Containment isolation within the time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a loss of coolant accident.
Additional valves have been identified as barrier valves, which in addition to the containment isolation valvo discussed above, are a part of the accident monitoring instrumentation in e chnical Specification 3/4.3.3.7 and are designated as Category 1 in au.ordance with Regulatory Guide 1.97, Revision 2,
" Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1980.
SEQUOYAH - UNIT 1 B 3/4 6-3 Amendment No. 67, 114, 150, 159, 203 1
CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES (Continued)
BASES l
The opening of penetration flow path (s) on an intermittent basis under administrative control includes the following considerations:
(1) stationing an operator, who is in constant communication with the control room, at the valve controls (2) instructing the operator to close these valves in an accident situation, and (3) assuring that the environmental conditions will not preclude access to close the valves and that this action will prevent the l
release of radioactivity outside the containment.
For valves with controls located in the control room, these conditions can be satisfied by including a specific reference to closing the particular valves in the emergency procedures, since communication and environmental factors are not affected because of the location of the valve controls.
SEQUOYAH - UNIT 1 B 3/4 6-3a Amendment No. 203
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UNITED STATES NUCLEAR REGULATORY COMMISSION C
WASHINGTON, D.C. 2056Mm01
\\,...../
t TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SE000YAH NUCLEAR PLANT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.193 License No. DPR-79 i
1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated April 6, 1995, 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; i
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.
i
. 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 Specifications l
The Technical Specifications contained in Appendices A and B, as revised through Amendment No.193, 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 60 days.
l FOR THE NUCLEAR REGULATORY COMMISSION Y
. IW Frederick J. Hebd n, Director Project Directorate II-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: June 13, 1995
ATTACHMENT TO LICENSE AMENDMENT NO.193 FACILITY OPERATING LICENSE N0. DPR-79 DOCKET N0. 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 l-2 1-2 3/4 6-1 3/4 6-1 3/4 6-2 3/4 6-2 3/4 6-4 3/4 6-4 3/4 6-5 3/4 6-6 1
3/4 6-6a 3/4 6-17 3/4 6-17 3/4 6-18 3/4 6-18 3/4 6-19 3/4 6-19 3/4 6-20 3/4 6-21 3/4 6-22 l
3/4 6-23 3/4 8-16 3/4 8-16 3/4 8-17 3/4 8-17 3/4 8-18 3/4 8-18 1
3/4 8-19 3/4 8-19 3/4 8-20 B3/4 6-2 B3/4 6-2 B3/4 6-3 B3/4 6-3 B3/4 6-3a r
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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 ser.sor as practicable to verify OPERABILITY I
including alarm and/or trip functions.
b.
Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.
c.
Digital, channels - the injection of a simulated signal into the chan-nel as close to the sensor input to the process racks as practicable to verify OPERABILITY including alarm and/or trip functions.
CfNTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:
a.
All penetrations required to be closed during accident conditions are either:
1)
Capable of being closed by an OPERABLE containment automatic isolation valve system, or 2)
Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are open under administrative control as permitted by Specifica-tion 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 Specifica-tion 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 Speci-fication 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 ALTERATIONS shall be the movement of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.
CORE OPERATING LIMIT REPORT 1.10 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document 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, 191, 193
3/4.6 CONTAINMENT SYSTEMS i
3/4.6.I PRIMARY CONTAINMENT CONTAINMENT INTECRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.
APPLICABILITY: MODES 1, 2, 3 and 4.
l 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 />.
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SURVEILLANCE REQUIREMENTS j
4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:
I a.
At least once per 31 days by verifying that all penetrations
- not capable of being closed by OPERABLE containment automatic isolation l
valves and required to be closed during accident conditions are closed by valves,' blind flanges, or deactivated automatic valves secured in their positions, except for valves that are open under administrative control as permitted by Specification 3.6.3.
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b.
By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3.
c.
Perform required visual examinations and leakage rate testing at P, 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.
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- Except valves, blind flanges, and deactivated automatic valves which are located inside the annulus or containment or the main steam valve vaults 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 i
need not be performed more often than once per 92 days.
SEQUOYAH - UNIT 2 3/4 6-1 Amendment No. 117, 167, 183, 193 i
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1 CONTAINMENT SYSTEMS SECONDARY CONTAINMENT BYPASS LEAKAGE 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
,penetrationsthataresecondarycontainmentBYPASSLEAKAGEPATHSTOTHE l
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 BTPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING, restore the combined bypass leakage rate from 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 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 />.
i SEQUOYAH - UNIT 2 3/4 6-2 Amendment No. 63, 167, 193
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I SEQUOYAH - UNIT 2 3/4 6-4 Amendment No. 63,90,104,117, 126, 167, 193 fM f v
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CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 Each containment isolation valve shall be OPERABLE.*
l APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
a.
With one or more of the isolation valve (s), except containment vacuum relief isolation valve (s), inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and either:
f 1.
Restore the inoperable valve (s) to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or 2.
Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one deactivated automatic valve secured in the isolation position, or 3.
Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual valve or blind flange; or 4.
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 />.
l b.
With one or more containment vacuum relief isolation valve (s) inoperable, l
l the valve (s) must be returned to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or be in i
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 l
the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
1 c.
The provisions of Specification 3.0.4 do not apply.
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SURVEILLANCE RE0VIREMENTS 4.6.3.1 Deleted l
l
- Penetration flow path (s) may be unisolated intermittently under administrative l
controls.
SEQUOYAH - UNIT 2 3/4 6-17 Amendment No.193
CONTAINMENT SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued) 4.6.3.2 Each automatic containment isolation valve shall be demonstrated l
OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by:
a.
Verifying that on a Phase A containment isolation test signal, each Phase A isolation valve actuates to its isolation position.
b.
Verifying that on a Phase B containment isolation test signal, each Phase B isolation valve actuates to its isolation position.
Verifying that on a Containment Ventilation isolation test signal, c.
each Containment Ventilation Isolation valve actuates to its isolation position.
d.
Verifying that on a high containment pressure isolation test signal, each containment vacuum relief valve actuates to its isolation position.
e.
Verifying that on a Safety Injection test signal that the Normal Charging Isolation valve actuates to its isolation position.
4.6.3.3 The isolation time of each power operated or automatic containment isolation valve shall be determined to be within its limit when tested pursuant to Specification 4.0.5.
l SEQUOYAH - UNIT 2 3/4 6-18 Amendment No. 72, 90, 104, 109, 193
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1 SEQUOYAH - UNIT 2
/ 6-19 Amendment No.193
e ELECTRICAL POWER SYSTEMS 3/4.8.3 ELECTRICAL E0VIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES LIMITING CONDITION FOR OPERATION 3.8.3.1 Primary and, Backup containment penetration conductor overcurrent protective devices associated with each containment electrical penetration circuit shall be OPERABLE. The scope of these protective devices excludes those circuits for which credible fault currents would not exceed the electrical penetration design rating.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With one or more of the containment penetration conductor overcurrent protective devices inoperable:
l a.
Restore the protective device (s) to OPERABLE status or de-energize the circuit (s) by tripping the associated backup circuit breaker within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and verify the backup circuit breaker to be tripped at least once per 7 days thereafter; the provisions of Specification 3.0.4 are not applicable to overcurrent devices in circuits which have their backup circuit breakers tripped, or b.
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 RE0VIREMENTS 4.8.3.1 All containment penetration conductor overcurrent protective device shall be demonstrated OPERABLE:
l a.
At least once per 18 months:
1.
For at least one 6.9 kV reactor coolant pump circuit, such that all reactor coolant pump circuits are demonstrated OPERABLE at least once per 72 months, by performance of:
(a) A CHANNEL CALIBRATION of the associated protective relays specified in appropriate plant instructions, and (b) An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits function as designed.
SEQUOYAH - UNIT 2 3/4 8-16 Amendment No. 34, 100, 193
ELECTRICAL POWER SYSTEMS l
SURVEILLANCE REQUIREMENTS (Continued)
(c) For each circuit breaker found inoperable during these functional tests, an additional representative sample of at least 1 of the circuit breakers of the inoperable type shall also be functionally l
l tested until no more failures are found or all circuit breakers of that, type have been functionally tested.
2.
By selecting and functionally testing a representative sample of at least 10% of each type of lower voltage circuit breakers. Circuit breakers selected for functional testing shall be selected on a rotating basis. The functional test shall consist of injecting a current input at the specified setpoint to each selected circuit breaker and verifying that each circuit breaker functions as designed. Circuit breakers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation.
For each circuit breaker found inoperable during these functional tests an additional representative sample of at least 10% of all the circuit breakers of the inoperable l
type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested.
3.
By selecting and verifying a representative sample of each type of fuse on a rotating basis. Verification will be accomplished as described by SR 4.8.3.1.a.3.a.
Each representative sample of fuses shall include at least 10% of all fuses of that type. Fuses found inoperable during l
verification shall be replaced with OPERABLE fuses prior to resuming operation.
For each fuse found inoperable during verification, an i
additional representative sample of at least 10% of all fuses of that type shall be functionally tested until no more failures are found or all fuses of that type have been functionally tested.
(a) A fuse verification and maintenance program will be maintained to ensure that:
1.
The proper size and type of fuse is installed, 2.
The fuse shows no sign of deterioration, and j
3.
The fuse connections are tight and clean, b.
At least once per 60 months by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedures prepared in conjunction with manufacturer's recomendations.
SEQUOYAH - UNIT 2 3/4 8-17 Amendment No. 34, 100, 193
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ELECTRICAL POWER SYSTEMS MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION LIMITING CONDITION FOR OPERATION l
3.8.3.2 The thermal overload protection devices, integral with the motor starter, of each valve used in safety systems shall be OPERABLE.
l APPLICABILITY: When'ever the motor operated valve is required to be OPERABLE.
f ACTION:
j With one or more of the thermal overload protection devices inoperable, declare i
the affected valve (s) inoperable and apply the ACTION Statement to the affected valve (s).
SURVEILLANCE REQUIREMENTS 4.8.3.2 The above required thermal overload protection devices shall be demonstrated OPERABLE:
a.
At least once per 18 months by the performance of a CHANNEL CALIBRATION of a representative sample of at least 25% of all thermal overload devices which are not bypassed, such that each non-bypassed device is calibrated at i
least once per 6 years, b.
At least once per 18 months, by the performance of a CHANNEL FUNCTIONAL TEST of the bypass circuitry for those thermal overload devices which are normally in force during plant operation and bypassed under accident conditions.
SEQUOYAH - UNIT 2 3/4 8-18 Amendment No. 25,71,115, 193
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SEQUOYAH - UNIT 2 3/4 8-19 Amendment No. 193 i
3/4.6 CONTAINMENT SYSTEMS BASES leakage paths to the auxiliary building is provided in plant procedures.
Restricting the leakage through the bypass leakage paths to 0.25 L provides assurance that the leakage fraction assumptions used in the evaluation 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 l
1eak rate.. Surveillance testing of the air lock seals provide assurance that t
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 pres:ure 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 STRL'CTURAL 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, 193
,fg-CONTAINMENT SYSTEMS BASES 3/4.6.1.8 EMERGENCY GAS TREATMENT SYSTEM (EGTS)
The OPERASILITY of the EGTS cleanup subsystem ensures that during LOCA l
conditions, containment vessel leakage into the annulus will be filtered through the HEPA filters and charcoal adsorber trains prior to discharge to the atmosphere. This requirement is necessary to meet the assumptions used in the accident analyses and limit the site boundary radiation doses to within the
. limits of 10 CFR 100 during LOCA conditions. Cumulative operation of the sys-ten with the heate'rs on 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 absorbers and HEPA filters. ANSI N510-1975 will be used as a procedural guide for surveillance testing.
3/4.6.1.9 CONTAINMENT VENTILATION SYSTEM Use of the containment purge lines is restricted to only one pair (one supply line and one exhaust line) of purge system lines at a time to ensure that the site boundary dose guidelines of 10 CFR Part 100 would not be exceeded in the event of a loss of coolant accident during purging operations. The analysis of this accident assumed purging through the largest pair of lines (a 24 inch inlet line and a 24 inch outlet line), a pre-existing iodine spike in the reactor coolant and four second valve closure times.
3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SUBSYSTEMS i
The OPERABILITY of the containment spray subsystems ensures that contain-ment depressurization and cooling capability will be available in the event of a LOCA. The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the accident analyses.
3/4.6.2.2 CONTAINMENT COOLING FANS The OPERABILITY of the lower containment vent coolers ensures that ade-quate heat removal capacity is available to provide long-term cooling following a non-LOCA event.
Postaccident use of these coolers ensures containment tem-peratures remain within environmental qualification limits for all safety-related equipment required to remain functional.
3/4.6.3 CONTAINMENT ISOLATION VALVES The operability of containment isolation valves ensures that the l
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. Containment isolation within the time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a loss of coolant accident.
Additional valves have been identified as barrier valves, which in addition to the containment isolation valves discussed above, are a part of the accident monitoring instrumentation in Technical Specification 3/4.3.3.7 and are desig-nated as Category 1 in accordance with Regulatory Guide 1.97, Revision 2, "Instrumeetation for Light-Water-cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1980.
SEQUOYAH - UNIT 2 B 3/4 6-3 Amendment No. 59,5 140, 19.
CONTAINMENT SYSTEMS BASES 3/4.6.3 CONTAINMENT ISOLATION VALVES (Continued)
The opening of penetration flow path (s) on an intermittent basis under administrative control includes the following considerations: (1) stationing an operator, who is in constant communication with the control room, at the valve controls, (2) instructing the operator to close these valves in an accident situation, and (3) assuring that the environmental conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside the containment.
For valves with controls located in the control room, these conditions can be satisfied by including a specific reference to closing the particular valves in the emergency procedures, since communication and environmental factors are not affected because of the location of the valve controls.
SEQUOYAH - UNIT 2 B 3/4 6-3a Amendment No.193
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