ML20216C116

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Amends 102 & 102 to Licenses NPF-37 & NPF-66,respectively, Deferring Next Scheduled Type a Containment Integrated Leak Rate Test for Plant,Unit 2,until Next Refueling Outage in 1999
ML20216C116
Person / Time
Site: Byron  Constellation icon.png
Issue date: 05/08/1998
From: John Hickman
Office of Nuclear Reactor Regulation
To:
Commonwealth Edison Co
Shared Package
ML20216C122 List:
References
NPF-37-A-102, NPF-66-A-102 NUDOCS 9805190106
Download: ML20216C116 (9)
v • d • e


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UNITED STATES

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NUCLEAR RESULATORY COMMISSION WASHINGTON, D.c. 2006dM201

          • ,o COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-454 BYRON STATION. UNIT NO.1 AMENDMENT TO FACfLITY OPERATING LICENSE Amendment No.102 License No. NPF-37 1.

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

A.

The application for amendment by Commonwealth Edison Company (the licensee) dated November 7,1997, as supplemented on March 24,1998, and April 9,1998, 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 provisi'ns 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.

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 Ucense No. NPF-37 is hereby amended to read as follows:

9805190106 980508 PDR ADOCK 05000454 P

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2 (2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No.102 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION dohn B. Hickman, Project Manager Project Directorate ill-2 Division of Reactor Projects - lil/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance: May 8, 1993 o

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,o COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-455 BYRON STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.102 License No. NPF-66 1.

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

A.

The application for amendment by Commonwealth Edison Company (the licensee) dated November 7,1997, as supplemented on March 24,1998, and April 9,1998, 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 j

(ii) that such activities will be conducted in compliance with the Commission's regulations, i

D.

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

i E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the I

Commission's regulations and all applicable requirements have been' satisfied.

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. NPF-66 is hereby amended to read as follows:

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. (2)

Technical Soecifications The Technical Specifications contained in Appendix A (NUREG-1113), as reviseo through Amendment No.102 and revised by Attachment 2 to NPF-66, and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF 37, dated February 14,1985, are hereby incorporated into this license. Attachment 2 contains a revision to Appendix A which is hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION I

n B. Hickman, Project Manager Project Directorate 111-2 Division of Reactor Projects - lil/lV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance:

May 8, 1998

ATTACHMENT TO LICENSE AMENDMENT NOS.102 AND 102 FACILITY OPERATING LICENSE NOS, NPF-37 AND NPF-66 DOCKET NOS. STN 50-454 AND STN 50-455 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marginallines indicating the area of change. Pages marked with an asterisk are provided for convenience.

Remove Paaes insert Paoes 3/4 6-1 3/4 6-1 3/4 & 2 3/4 & 2 3/4 6-3 3/4 6-3 8 3/4 6-1 B 3/4 6-1

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRIi /

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

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

ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 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 />.

SURVEll LANCE 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 3.6-1 of Specification 3.6.3 or for containment isolation valves that are open under administrative controls; b.

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

By performing containment leakage testing in accordance with Regulatory Guide 1.163, September 1995, as modified by an approved schedular exception and l

10 CFR 50, Appendix J, Option B.

  • Except valves, blind flanges, and deactivated automatic valves which are located inside the 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.

BYRON - UNITS 1 & 2 3/4 6-1 AMENDMENT NO. 102

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CONTAINMENT SYSTEM _S CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:

a.

An overall integrated leakage rate of less than or equal to L, at P..

b.

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

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

ACTIOfJ:

With either the measured overall integrated containment leakage rate exceeding 0.75 L, or the measured combined leakage rate for all penetrations and valves subject to Types B and C tests exceeding 0.60 L,, restore the overall integrated leakage rate to less than 0.75 L, and the combined leakage rate for all penetrations subject to Type B and C tests to less than 0.60 L.

prior to increasing the Reactor Coolant System temperature above 200 F.

SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated in accordance with Regulatory Guide 1.163, September 1995, and 10 CFR 50, Appendix J, Option B.

a.

Type A (Overall Integrated Containment Leakage Rate) testing shall be conducted in accordance with Regulatory Guide 1.163, September 1995, as modified by an approved schedular exception and 10 CFR 50, Appendix J, Option B.

BYRON - UNITS 1 & 2 3/4 6-2 AMENDMENT NO. 102

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CONTAINMENT SYSTEMS EURVEILLANCE REQUIREMENTS (Continued) b.

The reporting requirements and frequency of Type A tests shall be in accordance with Regulatory Guide 1.163, September 1995, as modified by an approved schedular exception and 10 CFR 50, Appendix J, Option B.

c.

The accuracy of each Type A test shall be verified by a supplemental test conducted in accordance with Regulatory Guide 1.163, September 1995, and 10 CFR 50, Appendix J, Option B.

l d.

Type B and C tests shai! be conducted in accordance with Regulatory Guide 1.163, September 1995, and 10 CFR 50, Appendix J, Option B.

e.

Air locks shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.3; f.

Purge supply and exhaust isolation valves with resilient material seals shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.7.3 or 4.6.1.7.4, as applicable; and l

g.

The structuralintegrity of the exposed accessible interior and exterior surfaces of the containment vessel, including the liner plate, shall be demonstrated during the shutdown for each Type A containment leakage rate test by a visualinspection of these surfaces. This inspection shall be performed at a frequency in accordance with Regulatory Guide 1.163, September 1995, to verify no apparent changes in l

appearance or other abnormal degradation.

l h.

The provisions of Specification 4.0.2 are not applicable.

BYRON - UNITS 1 & 2 3/4 6-3 AMENDMENT NO.102

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, will limit the SITE BOUNDARY radiation doses to within the dose guideline values of 10 CFR Part 100 during accident conditions.

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 accident analyses at the peak accident pressure, P,. As an added conservatism, the measured overallintegrated leakage rate is further limited to less than or equal to 0.75 L, during performance of the periodic test to account for possible degradation of the containment leakage barriers between leakage tests.

The surveillance testing for measuring leakage rates is consistent with the requirements of l

Appendix J of 10 CFR Part 50, Option B, Regulatory Guide 1.163, September 1995, Nuclear Energy Institute document NEl 94-01, and ANSI /ANS-56.8-1994 except as modified for Unit 2.

An extension for Unit 2 is allowed to perform the Type A test during B2RO8. Subsequent Type A test intervals for Unit 2 will be determined based on test results, in accordance with NEl 94-01.

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 provides 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 F'RESSURE The limitations on containment intemal pressure ensure that: (1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 0.1 psig, and (2) the containment peak pressure does not exceed the design pressure of 50 psig during steam line break conditions.

The maximum increase in peak pressure expected to be obtained from a cold leg double-onded break event is defined as P,. The limit of 1.0 psig for initial positive containment pressure will limit the total pressure to P,, which is higher than the UFSAR Chapter 15 accident analysis calculated peak pressure assuming a limit of 0.3 psig for initial positive containment pressure, but is considerably less than the design pressure of 50 psig.

BYRON - UNITS 1 & 2 B 3/4 6-1 AMENDMENT NO.102

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