ML20100J196

Amend 205 to License NPF-3,revising TSs 3/4.6.1.1,3/4.6.1.2, 3/4.6.1.6,& 3/4.6.5.3 & Associated Bases

ML20100J196
Person / Time
Site:	Davis Besse
Issue date:	02/22/1996
From:	Gundrum L NRC (Affiliation Not Assigned)
To:
Shared Package
ML20100J201	List: ML20100J196
References
NUDOCS 9602280320
Download: ML20100J196 (13)

Text

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4 KET UNITED STATES

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j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. enmaa anni

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TOLEDO EDISON COMPANY CENTERIOR SERVICE COMPANY ale THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DOCKET N0. 50-346 DAVIS-BESSE NUCLEAR POWER STATION. INIIT W. I AMENDMENT TO FACILITY OPERATING LICENSE i

Amendment No. 205 License No. NPF-3 1.

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

A.

'The application for amendment by the Toledo Edison Company,

.Centerior Service Company, and the Cleveland Electric Illuminating Company (the licensees) dated December 12, 1995, and supplemented by facsimile transmission dated January 26, 1996, 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 cosmon defense and security or to the health and safety of the public; and E.

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

2.

Accordingly, the license is amended by changes to the Technical

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Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-3 is hereby amended to read as follows:

9602280320 960222 ADOCK0500g6 DR

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Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.205, are hereby incorporated in the license.

The Toledo Edison Company shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented not later than 90 days after issuance.

FOR THE NUCLEAR REGULATORi L" 311SSION L

Linda L. Gundrum, Project Manager Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance:

February 22, 1996 1

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ATTACHMENT TO LICENSE AMENDMENT NO. 205 FACILITY OPERATING LICENSE NO. NPF-3 DOCKET NO. 50-346 i

Replace the following pages of the Appendix "A" Technical Specifications with j

the attached pages. The revised pages are identified by amendment numbes and

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contain vertical lines indicating the area of change.

I Remove Insert I

VI VI 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 1

3/4 6-9 3/4 6-9 l

3/4 6-32 3/4 6-32 B 3/4 6-1 B 3/4 6-1 B 3/4 6-2 B 3/4 6-2 B 3/4 6-5 B 3/4 6-5

MU LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION f1GE 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity...............

3/4 6-1 Containment Leakage................

3/4 6-2 Containment Air Locks...............

3/4 6-6 Internal Pressure.................

3/4 6-7 Air Temperature..................

3/4 6-8 Deleted......................

3/4 6-9 l

Containment Ventilation System 3/4 6-10 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System 3/4 6-11

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Containment Coeling System 3/4 6-13 3/4.6.3 CONTAINMENT ISOLATION VALVES 3/4 6-14 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Analyzers 3/4 6-23 4

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

3/4 6-24 Containment Hydrogen Dilution System 3/4 6-25 Hydrogen Purge System...............

3/4 6-26 4

i 3/4.6.5 SHIELD BUILDING Emergency Ventilation System 3/4 6-28 Shield Building Integrity.............

3/4 6-31 Deleted......................

3/4 6-32 l

l DAVIS-BESSE, UNIT I VI Amendment No. ??, SS, 135, 205 i

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d 3/4.6 CONTAlletENT SYSTEMS i

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

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

Without primary *,0NTAINMENT 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 i

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:

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All penetrations

• not capable of being closed by OPERABLE contain-ment 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 those valves that may be opened under administrative coritrols per Specification 3.6.3.I, and i

2.

All equipment hatches are closed and sealed.

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b. By verifying that each containment air lock is in compliance with the i

requirements of Specification 3.6.1.3 r

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c. By performing required visual examinations of the containment vessel and shield building in accordance with 10 CFR 50, Appendix J, Option i

l B, as modifie1 by approved exemptions, and in accordance with the l

guidelines contained in Regulatory Guide 1.163, dated September 1995.

• Except valves, blind flanges,'and deactivated automatic valves which are located inside the Shield Building (including the annulus and containment) and i

are locked, sealed, or otherwise secured in the closed position. These t

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

DAVIS-BESSE, UNIT 1 3/4 6-1 Amendment No. - 10, 194, 205 i.

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i CONTAIMENT SYSTEMS CONTAIMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2' Containment leakage rates shall be limited to:

a. An overall integrated leakage rate of < L,, 0.50 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at P,, 38 psig.

b. A combined leakage rate of < 0.60 L., for all penetrations and valves subject to Type B and C tests, when pressurized to P,.

c. A combined leakage rate of < 0.03 L, for all penetrations that are secondary containment bypass leakage paths, when pressurized to P.

i purge and exhaust isolation valve special tesl.for the containment l

d. A single penetration leakage rate of < 0.15 L APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

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a. With either: (a) the measured overall as-left integrated containment l leakage rate exceeding 0.75 L, (b) with the measured combined l

1eakage rate for all penetrations and valves subject to Type B and C tests exceeding 0.60 L, or (c) with the combined bypass leakage rate j

exceeding 0.03 L, resfore the leakage rate (s) to within the limit (s) l prior to increasing the Reactor Coolant System temperature above 200*F.

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b. With a single containment purge and exhaust isolation valve penetration having leakage rate exceeding 0.15 L ; restore the leakageratetowithinlimitsin72hoursorbeinatleastHOT 5

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 i

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

SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated as follows:

a. Perform Type A tests in accordance with 10 CFR 50, Appendix J, Option B, as modified by approved exemptions, and in accordance with the guidelines contained in Regulatory Guide 1.163, dated September, 1995.

DAVIS-BESSE, UNIT 1 3/4 6-2 Amendment No. 90, l'S, 150

-19th-205

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CONTAINMENT. SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued)

b. Deleted l

c. Deleted l

d. Perform Type B and C. tests in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions.

e. The combined bypass leakage rate shall be determined to be < 0.03 L, by applicable Type B and C tests at least once every 24 months except for penetrations which are not individually testable; penetrations not individually testable shall be determined to have no detectable leakage when tested with sotp bubbles while the containment is pressurized to P, 38 psig, during each Type A test.

f. Air locks shall be in compliance with the requirements of Specification 3/4.6.1.3.

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DAVIS-BESSE, UNIT 1 3/4 6-3 Amendment No. 120,150 7

49+r-i n, 205 h

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CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)

g. Each time the containment purge and exhaust isolation valves are opened, a special test shall be performed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after valve closure or prior to entering MODE 4 from MODE 5, whichever is later. The special test is conducted by pressurizing the piping section-including one valve inside and one valve outside the containment to a pressure greater or equal to 20 psig. The leakage rate per penetration shall not exceed 0.15 L,.

h. The special test as defined in Surveillance Requirement 4.6.1.2.g shall be perfomed for the containment purge and isolation valves when the plant has been in any combination of MODES 3, 4, 5 or 6 for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided that the test required by Surveillance Requirement 4.6.1.2.g has not been performed in the previous 6 months.

i. Deleted l

j. The provisions of Specification 4.0.2 are not applicable.

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DAVIS-BESSE, UNIT 1 3/4 6-4 Amendment No. 00, 100, 205 1

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Page 3/4 6-9 Deleted.

Next page is 3/4 6-10.

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' DAVIS-BESSE, UNIT 1 3/4 6-9

.hendment No. -937 205

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l Page 3/4 6-32 Deleted.

Next page is 3/4 6-33.

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DAVIS-BESSE, UNIT 1 3/4 6-32 Amendment No. -937 205

.. - _. _ _ q 3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3 /4. 6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation and air lock door i

requirements, will limit the site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions.

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3 /4. 6.1. 2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety conservatism, the measured overall as-left integrate 8. As an addedleakage rate is fur analyscs at the peak accident pressure of 38 psig, P limited to $ 0.75 L,, during performance of the periodic tests to account for possible degradation of the containment leakage barriers between leakage tests.

l The special test for the containment purge and exhaust isolation valves is intended to detect gross degradation of seals on the valve seats. The 1

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special test is performed in addition to the Appendix J requirements.

USAR 6.2.4 identifies all penetrations that are secondary containment bypass leakage paths.

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 1eak rate.

Surveillance testing of the air lock seals provide assurance that l

the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.

DAVIS-BESSE, UNIT 1 B 3/4 6-1 Amendment No. an, 345, 3pg

-104, 1;3, 205 i

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CONTAINMENT SYSTEMS BASES l

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 psi j

and 2) the containment peak pressure does not exceed the design pressure of 40 psig during LOCA conditions.

The maximum peak pressure obtained from a LOCA event is 37 psig.

The limit of I psig for initial positive containment pressure wn) limit I

the total pressure to 38 psig which is less than the design pressere and is consistent with the safety analyses.

3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the overall containment average air temperature does not exceed the initial temperature condition assumed in the accident analysis for a LOCA.

3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY 9eleted l

3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM The limitation on use of the Containment Purge and Exhaust System limits the time this system pay be in operation with the reactor coolant system temperature above 200 F.

This restriction minimizes the time that a direct open path would exist from the containment atmosphere to the outside atmosphere and consequently reduces the probability that an accident dose would exceed 10 CFR 100 guideline values in the event of a LOCA occurring coincident with purge system operation. The use of this system is therefore restricted to non-routine usage not to exceed 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> in any consecutive 365 day period which is equivalent to approximately 1% of the total possible yearly unit operating time.

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSlfM The OPERABILITY of the containment spray system ensures that containment depressurization and cooling capability will be available in the event of a LOCA. The pressure reduction and resultant lower containment DAVIS-BESSE, UNIT 1 B 3/4 6-2 Amendment No. 4Er 205

CONTAIM ENT SYSTEMS BASES 3/4.6.5.2 SHIELD BUILDING INTEGRITY SHIELDING BUILDING INTEGRITY ensures that the release of radioactive material from the cer.tainment vessel will be restricted to those leakage paths and associated leak rates assumed in the safety analysis. The closure of the airtight doors and blowout panels listed in Table 4.6-1 ensure that the Emergency Ventilation System (EVS) can provide a negative pressure between 0.25 and 1.5 inches Water Gauge within the annulus between the shield building and containment vessel and within the interconnecting mechanical penetration rooms after a loss-of-coolant accident (LOCA). This restriction, in conjunction with the operation of the EVS, will limit the site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions.

3/4.6.5.3 SHIELD BUILDING STRUCTURAL INTEGRITY Deleted l

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DAVIS-BESSE, UNIT 1 B 3/4 6-5 Amendment No. -95r205 4

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