Text

Amend 145 to License DPR-13,revising RETS in Accordance W/Generic Ltr 89-01,allowing Implementation of Programmatic Controls for RETS & REMP in AC Section of TS & Relocation of Procedural Details to ODCM or Pcp,As Appropriate
	ML13309A243
Person / Time
Site:	San Onofre
Issue date:	05/15/1992
From:	Quay T; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML13309A242	List: ML13309A240; ML13309A243; ML13309A244;
References
	DPR-13-A-145, GL-89-001 NUDOCS 9205290132
	Download: ML13309A243 (73)
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SjRREGIZq UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 20555 SOUTHERN CALIFORNIA EDISON COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY DOCKET NO. 50-206 SAN ONOFRE NUCLEAR GENERATING STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 145 License No. DPR-13

1.

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

A.

The applications for amendment by Southern California Edison Company (SCE or the licensee) and San Diego Gas and Electric Company dated February 26, 1988, and December 12, 1989, as clarified by letters dated July 27, 1990, and September 23, 1991, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, as amended, 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.

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 Provisional Operating License No. DPR-13 is hereby amended to read as follows:

9205290132 920515 PDR ADOCK 05000206 P

PDR

-2

2.

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 145, are hereby incorporated in the license.

Southern California Edison Company shall operate the facility in accordance with the Technical Specifications, except where otherwise stated in specific license conditions.

3.

This license amendment is effective as of the date of its issuance and must be fully implemented no later than 45 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

'Theodore R. Quay, Director Project Directorate V Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: May 15, 1992

ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 145 TO FACILITY OPERATING LICENSE NO. DPR-13 DOCKET NO. 50-206 Revise 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 areas of change.

REMOVE INSERT iv iv vi vi vii vii viii viii ix ix x

x xii xii xiii xiii xiv xiv 1.0-3 1.0-3 1.0-4 1.0-4 1.0-5 1.0-5 3.5-16 3.5-16 3.5-26 through 3.5-34 3.5-26 through 3.5-30 3.15-1 3.15-1 3.15-2 3.15-3 3.16-1 through 3.16-7 3.16-1 through 3.16-6 3.17-1 3.17-1 3.17-2 3.18-1 through 3.18-9 3.18-1 3.19-1 3.19-1 4.1-13 through 4.1-35 4.1-13 through 4.1-31 4.5-1 through 4.5-6 4.5-1 4.6-1 through 4.6-9 4.6-1 through 4.6-6 4.17-1 4.17-1 4.18-1 through 4.18-6 4.18-1 4.19-1 4.19-1 6.5-4 6.5-4 6.8-3 6.8-3 6.8-4 6.8-5 6.9-2 6.9-2 6.9-3 6.9-3 6.9-4 6.9-5 6.10-2 6.10-2 6.13-1 6.13-1 6.14-1 6.14-1 6.15-1 6.15-1

APPENDIX A TECHNICAL SPECIFICATIONS LIST OF EFFECTIVE PAGES Page Amendment No.

Page Amendment No.

i 143, 145 2.1-4 43, 117, 121, 122, 3.1-23 83, 130 ii 143, 145 130 3.2-1 102, 130 iii 143, 1L 2.1-5 43, 97, 117, 121, 3.2-2 25, 102, 130 iv 139, la 122, 130 3.3-1 25, 37, 86, 124 v

90, 130, 131 2.1-6 55, 130 130 vi 90, 130, 131, 45 3.0-1 43, 56, 64, 83 3.3-2 25, 130 vii 90, 102, 130, 131 130 3.3-3 25, 38, 86, 124, 3.0-2 56, 64, 83, 130, 130, 142 viii 90, 130, 131, 145 135 3.3-4 25, 37, 124, 130 ix 90, 91, 102, 130 3.0-3 135 3.3-5 25, 130 131, 145 3.0-4 135 3.3-6 25, 102, 120, 130 x

90, 91, 130, 131 3.1-1 29, 38, 70, 83, 3.3-7 25, 102, 130 Li45 91, 96, 130 3.3-8 25, 38, 122, 130 xi 55, 92, 102, 110, 3.1-2 29, 83, 96, 130 3.3-9 25, 38, 122, 130 111, 130, 131, 143 3.1-3 43, 77, 103, 130 3.3-10 25, 130 xii 56, 58, 71, 79, 3.1-4 77, 130 3.3-11 NRC Order 83, 104, 117, 130, 3.1-5 77, 125, 130 4/20/81, 130 131, 143, 3.1-6 77, 102, 130 3.3-12 NRC Order xiii 31, 56, 3.1-7 77, 102, 103, 130 4/20/81, 130 83, 84, 91, 117, 3.1-8 43, 102, 103, 130 3.4-1 29, 82, 125, 130 130, 131, 143, 1_5 3.1-9 77, 102, 130 3.4-2 29, 130 xiv 131, 143, L45 3.1-10 Change No. 14, 38, 3.4-3 82, 125, 130, 138 xv 143 102, 130 3.4-4 82, 125, 130, 138 1.0-1 31, 56, 59, 83, 3.1-11 Change No. 14, 38, 3.5-1 83, 117, 130, 143 117, 130 102, 130 3.5-2 43, 56, 58, 83, 1.0-2 31, 56, 59, 83, 3.1-12 Change No. 14, 92, 117, 128, 130 104, 117, 130 130 3.5-3 43, 56, 58, 83, 1.0-3 31, 56, 59, 79, 3.1-13 Change No. 14, 92,

117, 121,

122, 83, 104, 117, 130, 130 130, 143 145 3.1-14 Change No. 14, 3.5-4 55, 58, 83, 117, 1.0-4 31, 56, 59, 79, 102, 130 118, 121, 128, 130 83, 117, 130, 145 3.1-15 Change No. 14, 3.5-5 83, 117, 130 1.0-5 77, 79, 83, 1177, 102, 130 3.5-5a 143 130, 145 3.1-16 Change No. 14, 3.5-5b 143 1.0-6 79, 85,796, 117, 102, 130 3.5-Sc 143 130 3.1-17 Change No. 14, 3.5-6 7, 11, 25, 35, 55, 1.0-7 58, 83, 117, 130 102, 130 56, 111, 130 1.0-8 56, 83, 117, 130 3.1-18 Change No. 7, 37, 3.5-7 7, 11, 25, 35, 49, 2.1-1 43, 55, 97, 117, 55, 91, 119, 130 55, 56, 111, 122, 130 3.1-19 Change No. 7, 55, 130 2.1-2 43, 97, 117, 121, 119, 130 3.5-8 11, 49, 111, 122, 130 3.1-20 37, 55, 119, 130 130 2.1-3 43, 117, 121, 122, 3.1-21 58, 59, 83, 130 3.5-9 11, 25, 56, 111, 130 3.1-22 58, 130 130 SAN ONOFRE -

UNIT 1 i

AMENDMENT NO. 145

LIST OF EFFECTIVE PAGES (Cont.)

Page Amendment No.

3.5-10 56, 130 3.7-7 134, 136 3.14-24 130, 131 3.5-11 56, 130 3.7-8 136 3.14-25 130, 131 3.5-12 56, 130 3.8-1 25, 36, 37, 3.14-26 130, 131 3.5-13 56, 130 73, 77, 130 3.15-1 79, 130,145 3.5-14 56, 130 3.8-2 36, 77, 130 3.16-1 79, 130, 145 3.5-15 58, 83, 130 3.8-3 25, 36, 73, 3.16-2 79, 90, 91 3.5-16 58, 83, 130, 145 77, 130 105, 130, 145 3.5-17 58, 83, 130 3.8-4 36, 73, 77, 130 3.16-3 79, 90 91, 3.5-18 58, 72, 130 3.8-5 36, 73, 77, 130 105, 130,145 3.5-19 58, 83, 125, 130 3.9-1 3, 10, 112, 130 3.16-4 79, 90, 91, 3.5-20 125, 130 3.9-2 3, 10, 112, 130 105, 130,145 3.5-21 58, 83, 117, 124, 3.10-1 7, 8, 112, 122, 3.16-5 79, 90, 137, 125, 130 130 145 3.5-22 64, 82, 125, 130 3.10-2 7, 112, 122, 130 3.16-6 7, 90, 130, 3.5-23 58, 82, 125, 130 3.11-1 7, 8, 11, 35, 145 3.5-24 58, 82, 125, 130 55, 112, 117, 3.17-1 7, 90, 91, 3.5-25 58, 82, 125, 129, 122, 130 105, 130,145 130 3.11-2 7, 11, 112, 3.18-1 79, 91, 1057,130, 3.5-26 79, 90, 105, 130 117, 122, 130 145 145 3.12-1 14, 130 3.19-1 7F, 90, 105, 130, 3.5-27 79, 90, 105, 130 3.13-1 21, 63, 81, 130 145 145 3.13-2 21, 63, 81, 130 3.20-1 In, 130 3.5-28 83, 130, 145 3.14-1 31, 44, 93, 130, 3.20-2 102, 130 3.5-29 83, 130, 145 131 4.0-1 83, 130, 137 3.5-30 83, 91, 130, 145 3.14-2 31, 130, 131 4.0-2 83, 130, 137 3.6-1 25, 56, 58, 3.14-3 31, 130, 131 4.0-3 135 73, 130 3.14-4 31, 93, 130, 131 4.0-4 135 3.6-2 Change No. 7, 3.14-5 31, 130, 131 4.1-1 29, 56, 83, 56, 130 3.14-6 130, 131, 134 117, 130 3.6-3 58, 99, 130 3.14-7 130, 131 4.1-la 143 3.6-4 58, 71, 99, 130 3.14-8 130, 131 4.1-2 7, 22, 83, 117, 3.6-5 58, 59, 71, 3.14-9 130, 131 122, 130, 143 99, 130 3.14-10 130, 131 4.1-3 117, 130 3.6-6 58, 59, 83, 130 3.14-11 31, 93, 130, 131 4.1-4 25, 29, 70 3.7-1 25, 52, 68, 84, 3.14-12 31, 130, 131 96, 117, 130 130, 134, 136 3.14-13 31, 130, 131 4.1-5 25, 29, 70, 3.7-2 25, 84, 106, 130, 3.14-14 31, 130, 131 96, 130 134, 136 3.14-15 31, 93, 130, 131 4.1-6 25, 29, 56, 70, 3.7-3 25, 84, 130, 134, 3.14-16 31, 130, 131 100, 122, 130, 134 136 3.14-17 31, 130, 131, 134 4.1-7 25, 29, 36 3.7-4 25, 52, 68, 84, 3.14-18 31, 130, 131 77, 103, 130, 142 106, 130, 134, 136 3.14-19 130, 131 4.1-8 77, 125, 130 3.7-5 25, 52, 68, 84, 3.14-20 130, 131 4.1-9 77, 122, 125, 130 106, 120, 130, 3.14-21 130, 131 134, 136, 143 3.14-22 130, 131

.7-6 134, 136, 143 3.14-23 130, 131 SAN ONOFRE - UNIT 1 ii AMENDMENT NO.145

LIST OF EFFECTIVE PAGES (Cont.)

Page Amendment No.

4.1-10 117, 130 4.3-6 75, 87, 130 4.14-4 21, 63, 81, 105, 4.1-11 25, 130 4.3-7 58, 130 130 4.1-12 Change No. 5, 25, 4.3-8 58, 59, 83, 130 4.15-1 31, 37, 38, 44, 130 4.3-9 58, 83, 130 130, 131 4.1-13 79, 130, 145 4.4-1 Change No. 12, 25, 4.15-2 31, 44, 130, 131 4.1-14 79, 130, 145 56, 82, 84, 95, 4.15-3 31, 130, 131 4.1-15 58, 130, 145 104, 123, 130, 136 4.15-4 130, 131 4.1-16 58, 130, 145 4.4-2 25, 84, 130, 136 4.15-5 130, 131 4.1-17 58, 130, 145 4.4-3 25, 56, 82, 84, 4.15-6 31, 130, 131 4.1-18 58, 83, 155, 145 117, 130, 134, 136 4.15-7 31, 130, 131 4.1-19 58, 83, 117, III, 4.4-4 25, 82, 84, 95, 4.15-8 130, 131 130, 145 104, 123, 124, 4.15-9 130, 131 4.1-20 58, lW, 130, 145 130, 134, 136, 143 4.15-10 130, 131 4.1-21 58, 130, 145 4.4-5 84, 130, 136 4.15-11 130, 131 4.1-22 58, 82, 126, 130, 4.4-6 25, 56, 84, 104, 4.15-12 130, 131 123, 130, 136 4.16-1 31, 101, 115, 130, 4.1-23 58, 82, 125, 130, 4.4-7 25, 56, 84, 130, 133 145 134, 136 4.16-2 37, 101, 115, 130, 4.1-24 E, 82, 125, 130, 4.4-8 136 133 145 4.4-9 136 4.16-3 37, 101, 115, 130,

.1-25 M, 125,

130, 4.4-10 136 133 Revised by NRC 4.5-1 Change No. 10, 38, 4.16-4 37, 55, 101, 115, letter 2/12/92 145 79, 130,145 130, 133 4.1-26 82, 130, 145 4.6-1 38, 79, 1

,145 4.16-5 37, 91, 101, 105, 4.1-27 83, 130,145 4.6-2 79, 130,145 115, 130, 133 4.1-28 83, 130, 145 4.6-3 79, 130, f47 4.17-1 79, 130, 145 4.1-29 83, 130, 145 4.6-4 79, 130, f4T 4.18-1 79, 130, 145 4.1-30 119, 130, 1A1 4.6-5 79, 130, 145 4.19-1 79, 105, ID, 145 4.1-31 119, 130, 145 4.6-6 79, 130, fT5 4.20-1 102, 130 4.2-1 25, 37, 54, 114, 4.7-1 Change No.14, 46, 5.1-1 25, 72, 130 130 130 5.1.2 72, 79, 130 4.2-2 25, 37, 114, 130 4.8-1 91, 130 5.2-1 72, 130 4.2-3 25, 37, 114, 130 4.9-1 Change No. 14, 130 5.2-2 72, 130 4.2-4 25, 37, 114, 124, 4.9-2 130 5.2-3 72, 130 130 4.10-1 13, 94, 130 5.3-1 3, 72, 130 4.2-5 25, 37, 114, 124, 4.10-2 13, 94, 130 5.3.2 3, 37, 72, 130 130 4.11-1 14, 109, 130 5.4-1 130 4.2-6 37, 114, 130 4.11-2 14, 109, 130 6.1-1 12, 39, 66 4.2-7 NRC Order 4.12-1 91, 130, Revised 91, 130 4/20/81, 54, 130 by NRC letter 6.2-1 44, 66, 91 4.3-1 24, 58, 75, 130 2/12/92 110, 130 4.3-2 5, 24, 58, 75, 87, 4.13-1 18, 113, 6.2-2 88, 91, 130 118, 130 130 6.2-3 66, 88' 91, 4.3-3 24, 58, 87, 118, 4.14-1 21, 33, 63, 81, 105, 110, 126, 130 130 130

.3-4 24, 87, 130 4.14-2 21, 63, 81, 130 6.2-4 12, 58, 91, 130

.3-5 24, 75, 87, 130 4.14-3 21, 63, 81, 130 SAN ONOFRE -

UNIT 1 iii AMENDMENT NO. 145

LIST OF EFFECTIVE PAGES (Cont.)

Page Amendment No.

6.3-1 12, 27, 39 6.13-1 58, 79, 91, 130, 54, 58, 66, 145 91, 130 6.14-1 3E, 79, 91, 130, 6.4-1 31, 39, 54, 145 91, 130, 139 6.15-1 NE, 79, 91, 130, 6.5-1 Change No. 69, 145 12, 39, 42, 6.16-1 5, 91, 130 54, 66, 91, 130 6.16-2 90, 91, 130 6.5-2 39, 54, 66, 6.16-3 90, 91, 105, 130 91, 130 6.16-4 90, 91, 130 6.5-3 12, 16, 39, 50, 54, 66, 91, 130 6.5-4 39, 66, 69, 79, 91, 130, jn 6.5-5 12, 50, 90, 91, 130 6.5-6 12, 16, 69, 91, 105, 130 6.5-7 12, 31, 91, 110, 130 6.6-1 12, 91, 130 6.7-1 39, 54, 66, 91, 130 6.8-1 39, 54,.66, 79, 91, 105, 130 6.8-2 39, 91, 130 6.8-3 91, 130, 145 6.8-4 145 6.8-5 155 6.9-1 12, 16, 30, 91, 130 6.9-2 12, 16, 30, 79, 91, 96, 130 145 6.9-3 12, 15, 16, 30, 79, 91, 96, 105 130, 145 6.10-1 12, 15T,16, 91, 130 6.10-2 12, 63, 81, 91, 130, 4j 6.11-1 58, 61, 91, 130 0

6.12-1 23, 38, 58, 91, 130 6.12-2 23, 38, 91, 130 SAN ONOFRE -

UNIT 1 iv AMENDMENT NO. 145

SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS SECTION PAGE 3.4 TURBINE CYCLE........................

3-4-1 3.4.1 Operating Status...

3.4-1 3.4.2 Maximum Secondary Coolant Activity......

3.4-3 3.4.3 Auxiliary Feedwater System..

3.4-4 3.4.4 Auxiliary Feedwater Storage Tank........

3.4-5 3.5 INSTRUMENTATION AND CONTROL................

3.5-1 3.5.1 Reactor Trip System Instrumentation.......

3.5-1 3.5.2 Control Rod Insertion Limits.

3.5-6 3.5.3 Control and Shutdown Rod Misalignment.*.'.*.*.....

3.5-10 3.5.4 Rod Position Indicating System......

3.5-13 3.5.5 Containment Isolation Instrumentation.....

3.5-15 3.5.6 Accident Monitoring Instrumentation....

3.5-19 3.5.7 Auxiliary Feedwater Instrumentation.....

3.5-22 3.5.8 [Deleted].

3.5-26 3.5.9 [Deleted].

3.5-27 3.5.10 Radiation Monitoring Instrumentation...

3.5-28 3.6 CONTAINMENT SYSTEMS........

3.6-1 3.6.1 Containment Sphere...

3.6-1 3.6.2 Containment Isolation Valves.

3.6-3 3.6.3 Hydrogen Monitors and Hydrogen Recombiners.....

3.6-6 3.7 AUXILIARY ELECTRICAL SUPPLY.....

3.7-1 3.8 FUEL LOADING AND REFUELING.............

3.8-1 3.9 MODERATOR TEMPERATURE COEFFICIENT (MTC)......

3.9-1 3.10 INCORE INSTRUMENTATION.

..3.10-1 3.11 CONTINUOUS POWER DISTRIBUTION MONITORING......

3.11-1 3.12 CONTROL ROOM EMERGENCY AIR TREATMENT SYSTEM......

3.12-1 3.13 SHOCK SUPPRESSORS (SNUBBERS) OPERABILITY....

3.13-1 3.14 FIRE PROTECTION SYSTEMS OPERABILITY......

3.14-1 3.14.1 Fire Suppression Water System.........

3.14-1 3.14.2 Spray and/or Sprinkler Systems.......

...3.14-3 3.14.3 Foam Suppression System..........

3.14-7 3.14.4 Halon Systems................

3.14-9 3.14.5 Fire Hose Stations..............

3.14-11 SAN ONOFRE -

UNIT 1 vi AMENDMENT NO: 00, 120, 131, 145

SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS SECTION PAGE 3.14.6 Fire Detection Instrumentation......

3.14-15 3.14.7 Fire Barriers.....

3.14-19 3.14.8 Dedicated and Alternate Shutdown Systems.

3.14-21 3.14.9 Eight Hour Emergency Lighting Units....

3.14-26 3.15 [DELETED]..

3.15-1 3.16 RADIOACTIVE GASEOUS EFFLUENTS..............

3.16-1 3.16.1 [Deleted]..

3.16-1 3.16.2 [Deleted]..

3.16-2 3.16.3 [Deleted]...

3.16-3 3.16.4 [Deleted]...

3.16-4 3.16.5 Gas Storage Tank.......

3.16-5 3.16.6 Explosive Gas Mixture.....

3.16-6 3.17 [DELETED]........

3.17-1 3.18 [DELETED]...

3.18-1 3.19 [DELETED]....

3.19-1 3.20 OVERPRESSURE PROTECTION SYSTEMS.......

3.20-1 SECTION 4 SURVEILLANCE REQUIREMENTS 4.0 SURVEILLANCE REQUIREMENTS (GENERAL)..

4.0-1 4.1.1 Operational Safety Items..............4.1-1 4.1.2 [Deleted]......................

4.1-13 4.1.3 [Deleted)...........................

4.1-14 4.1.4 Containment Isolation Instrumentation........

4.1-15 4.1.5 Accident Monitoring Instrumentation........4.1-18 4.1.6 Pressurizer Relief Valves.............4.1-20 4.1.7 Pressurizer............4.1-21 4.1.8 Auxiliary Feedwater Instrumentation.........

4.1-22 4.1.9 Auxiliary Feedwater System Surveillance.......

4.1-24 4.1.10 AuxiliaryFeedwater Storage Tank Surveillance.

4.1-26 4.1.11 Radiation Monitoring Instrumentation........4.1-27 4.1.12 Reactor Coolant System Vents...

4.1-29 41.13 Leakage and Leakage Detection Systems......

4.1-30 4.2 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEM.........

4.2-1 4.2.1 Safety Injection and Containment Spray..

4.2-1 System Periodic Testing 4.2.2 Primary Coolant System Pressure.......

4.2-7 Isolation Valves Testing (Surveillance Requi rement) 44.2.1 SA NOR

-UIT1vi MEDET4O.2.20,flIZ 4

SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS SECTION PAGE 4.3 CONTAINMENT SYSTEMS..

4.3-1 4.3.1 Containment Testing.

4.3-1 4.3.2 Containment Isolation Valves........

......4.3-7 4.3.3 Hydrogen Monitors and Hydrogen Recombiners......

4.3-8 4.4 EMERGENCY POWER SYSTEM PERIODIC TESTING.

4.4-1 4.5

[DELETED]...

4.5-1 4.6 RADIOACTIVE GASEOUS EFFLUENTS....

4.6-1 4.6.1 [Deleted]..

4.6-1 4.6.2 [Deleted]...

4.6-2 4.6.3 [Deleted].......

4.6-3 4.6.4 [Deleted].....

4.6-4 4.6.5 Gas Storage Tank......

4.6-5 4.6.6 Explosive Gas Mixture....

4.6-6 4.7 INSERVICE INSPECTION REQUIREMENTS...

4.7-1 4.8 REACTIVITY ANOMALIES........

4.8-1 4.9 REACTOR VESSEL SURVEILLANCE PROGRAM.......

.....4.9-1 4.10 AUGMENTED INSERVICE INSPECTION OF HIGH........

4.10-1 ENERGY LINES OUTSIDE CONTAINMENT 4.11 CONTROL ROOM EMERGENCY AIR TREATMENT SYSTEM......

4.11-1 4.12 MISCELLANEOUS RADIOACTIVE MATERIAL SOURCES...

4.12-1 4.13 [DELETED].

..4.13-1 4.14 SHOCK SUPPRESSORS (SNUBBERS) SURVEILLANCE.......

4.14-1 4.15 FIRE PROTECTION SYSTEMS SURVEILLANCE.

4.15-1 4.15.1 Fire Suppression Water System............4.15-1 4.15.2 Spray and/or Sprinkler Systems...........4.15-3 4.15.3 Foam Suppression Systems..............4.15-4 4.15.4 Halon System..................4.15-5 4.15.5 Fire Hose Stations.................

4.15-6 4.15.6 Fire Detection Instrumentation...........

4.15-7 4.15.7 Fire Barriers...

S...s.............. 4.15-8 4.15.8 Dedicated and Alternate Shutdown Systems.

4.15-9 Surveillance 4.15.9 Eight Hour Emergency Lighting Units Surveillance 4.15-12 SAN ONOFRE -

UNIT 1 viii AMENDMENT NO: 00, DO0, Z, 145

SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 1 TECHNICAL SPECIFICATIONS SECTION TABLE OF CONTENTS PAGE 4.16 INSERVICE INSPECTION OF STEAM GENREATOR TUBING...

4.16-1 4.17 [DELETED]......

........ 4.17-1 4.18 [DELETED].......

4.18-1 4.19 [DELETED]......

4.19-1 4.20 OVERPRESSURE PROTECTION SYSTEMS...............

4.20-1 SECTION 5 DESIGN FEATURES 5.1 SITE DESCRIPTION......

5.1-1 5.2 CONTAINMENT..... * *. * * * * *....

5.2-1 5.3 REACTOR......

5.3-1 5.4 AUXILIARY EQUIPMENT.

5.4-1 SECTION 6 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY.

6.1-1 6.2 ORGANIZATION.....

6.2-1 6.2.1 Offsite and Onsite Organization....

6.2-1 6.2.2 Unit Staff.........

6.2-1 6.3 UNIT STAFF QUALIFICATIONS..

6.3-1 6.4 TRAINING......

6.4-1 6.5 REVIEW AND AUDIT.

6.5-1 6.5.1 Onsite Review Committee (OSRC)..

6.5-1 6.5.2 Technical Review and Control.........*....

6.5-3 6.5.3 Nuclear Safety Group (NSG)..........

6.5-5 6.6 REPORTABLE EVENT ACTION.....

6.6-1 6.7 SAFETY LIMIT VIOLATION.....*...

6.7-1 6.8 PROCEDURES AND PROGRAMS........

6.8-1 SAN ONOFRE -

UNIT 1 ix AMENDMENT NO:

Z, SI, 202, Z30, 1Z, 145

SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS SECTION PAGE 6.9 REPORTING REQUIREMENTS.................

6.9-1 6.9.1 Routine Reports...........

6.9-1 6.9.2 Special Reports..

6.9-3 6.10 RECORD RETENTION.

..6.10-1 6.11 RADIATION PROTECTION PROGRAM....

.......6.11-1 6.12 HIGH RADIATION AREA.

6.12-1 6.13 PROCESS CONTROL PROGRAM (PCP)...

6.13-1 6.14 OFFSITE DOSE CALCULATION MANUAL (00CM)...

6.14-1 6.15 [DELETED]..

6.15-1 6.16 ENVIRONMENTAL PROTECTION........

6.16-1 SAN ONOFRE -

UNIT 1 x

AMENDMENT NO:

00, SI, 130, 1ZZ, 145

SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 1, TECHNICAL SPECIFICATIONS TABLES TABLES TITLE PAGE 1.1 Frequency Notation 1.0-7 1.2 Operational Modes 1.0-8 2.1 Maximum Safety Systems Settings 2.1-5 3.1.3.1 San Onofre Nuclear Generating Station Unit 1 3.1-16 Reactor Vessel Toughness Data (unirradiated) 3.3.5-1 Primary Coolant System Pressure Isolation Valves 3.3-12 3.5.1-1 Reactor Trip System Instrumentation 3.5-2 3.5.1-2 Reactor Trip System Instrumentation Trip Setpoints 3.5-5a 3.5.1-3 Reactor Trip System Instrumentation Response Times 3.5-5b 3.5.3-1 Accident Analyses Requiring Reevaluation -in 3.5-12 the Event of an Inoperable Rod 3.5.5-1 Containment Isolation Instrumentation 3.5-16 3.5.5-2 Containment Isolation Instrumentation Trip 3.5-18 Set Points 3.5.6-1 Accident Monitoring Instrumentation 3.5-21 3.5.7-1 Auxiliary Feedwater Instrumentation 3.5-23 3.5.7-2 Auxiliary Feedwater Instrumentation Trip 3.5-25 Set Points 3.5.8-1

[Deleted]

3.5.9-1

[Deleted]

3.5.10-1 Radiation Monitoring Instrumentation 3.5-33 3.6.2-1 Power Operated or Automatic Containment Isolation 3.6-5 Valve Summary 3.14.2-1 Required Sprinkler and Spray Systems 3.14-6 3.14.5.1 Fire Hose Stations 3.14-13 SAN ONOFRE - UNIT 1 xii AMENDMENT NO: %X, 0, 58, 70 OX, 84, 01, 117 130,131, 143, 145

SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 1 TECHNICAL SPECIFICATIONS TABLES TABLES TITLE PAGE 3.14.6.1 Fire Detection Instruments 3.14-17 3.14.8.1 DSD and Alternate Shutdown Minimum Operable Equipment 3.14-24 3.18.1

[Deleted]

3.18.2

[Deleted]

4.1.1 Reactor Trip System Instrumentation Surveillance 4.1-2 Requirements 4.1.2 Minimum Equipment Check and Sampling Frequency 4.1-4 4.1.3 Minimum Frequencies for Testing, Calibrating, 4.1-10 and/or Checking of Instrument Channels 4.1.2.1

[Deleted]

4.1.3.1

[Deleted]

4.1.4-1 Containment Isolation Instrumentation 4.1-16 Surveillance Requirements 4.1.5-1 Accident Monitoring Instrumentation 4.1-19 Surveillance Requirements 4.1.8-1 Auxiliary Feedwater Instrumentation 4.1-23 Surveillance Requirements 4.1.11-1 Radiation Monitoring Instrumentation 4.1-28 Surveillance Requirements 4.1.13-1 Leakage Detection Systems 4.1-31 4.4-1 Battery Surveillance Requirements 4.4-5 4.5.1.1

[Deleted]

4.6.1.1

[Deleted]

SAN ONOFRE -

UNIT 1 xiii AMENDMENT NO: SI, 50, 50, 70 8V, 84, 97, 117 130, 13I, W43, 145

SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 1 TECHNICAL SPECIFICATIONS TABLES TABLES TITLE PAGE 4.18.1

[Deleted]

6.2-1 Minimum Shift Crew Composition 6.2-4 SAN ONOFRE -

UNIT 1 xiv AMENDMENT NO:

141, 14.5

O E - AVERAGE DISINTEGRATION ENERGY E is the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines and tritium with half lives greater than 15-minutes, making up at least 95% of the total non-iodine and non-tritium activity in the coolant.

FIRE SUPPRESSION WATER SYSTEM A FIRE SUPPPRESSION WATER SYSTEM shall consist of a water source(s), pump(s),

and distribution piping with associated isolation valves (i.e., system header, hose standpipe and spray header isolation valves).

FREQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

GASEOUS RADWASTE TREATMENT SYSTEM A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgas from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

MEMBER(S) OF THE PUBLIC MEMBER(S) of THE PUBLIC shall include all individuals who by virtue of their occupational status have no formal association with the plant. This category shall include non-employees of the licensee who are permitted to use portions of the site for recreational, occupational, or purposes not associated with plant functions. This category shall not, include non-employees such as vending machine servicemen or postmen who, as part of their formal job function, occasionally enter an area that is controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.

OFFSITE DOSE CALCULATION MANUAL (ODCM)

The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Environmental Radiological Effluent Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Semiannual Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.8.

SAN ONOFRE -

UNIT 1 1.0-3 AMENDMENT NO:

31, SE, BO, 70 OZ, 104, 117, 110,145

OPERABLE - OPERABILITY A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s), and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE - MODE An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.

PROCESS CONTROL PROGRAM The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

. PURGE-PURGING PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

RATED THERMAL POWER RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 1347 Mwt.

REPORTABLE EVENT A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

RESIDUAL HEAT REMOVAL (RHR) TRAIN An RHR TRAIN shall be a train of components that includes: one RHR pump aligned with one RHR heat exchanger; one component cooling water pump aligned with the same RHR heat exchanger and with one component cooling water heat exchanger; and one salt water pump aligned with the same component cooling water heat exchanger.

SAN ONOFRE - UNIT 1 1.0-4 AMENDMENT NO:

X1, 50, $0, 70 OX, III, 120.145

SHUTDOWN MARGIN SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee.

SOLIDIFICATION

[Deleted]

SOURCE CHECK A SOURCE CHECK is the qualitative assessment of a channel response when the channel sensor is exposed to a radioactive source.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of:

a. A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals,

b. the testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TRIP ACTUATING DEVICE OPERATIONAL TEST A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary, of the Trip Actuating Device such that it actuates at the required setpoint within the required accuracy.

SAN ONOFRE - UNIT 1 1.0-5 AMENDMENT NO:

77, 70, $Z, 117, 130, 145

TAB L

.5.5-1 CONTAINMENT ISOLATION INSTRUMENTATION C>

MINIMUM ITOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION Containment Isolation (Valves listed In Table 3.6.2-1) a) Manual 2

1 2

1, 2, 3, 4 11 b) Containment Pressure-High 3/train 2/train 2/train 1, 2, 3 9

c) Sequencer Subchannels 2/sequencer 2/sequencer 2/sequencer 1, 2, 3, 4 8

d) Safety Injection

1) Containment Pressure-High 3/train 2/train 2/train 1, 2, 3 9*

2) Pressurizer Pressure-Low 3/train 2/train 2/train 1, 2, 3 9*

Purge and Exhaust Isolation (POV-9, POV-10, CV-10, CV-40, CV-116) a) Manual 1/valve 1/valve 1/valve 1, 2, 3, 4 10 r

b) Containment 1

Radioactivity-High 1

1 1,2,3,4 10*

Mr

-4 0Q Cii

3.5.8 [Deleted]

SAN ONOFRE - UNIT 1 3.5-26 AMENDMENT NO: 79, 00, 109, ZZ0, 145

O 3.5.9 [Deleted]

SAN ONOFRE -

UNIT 1 3.5-27 AMENDMENT NO: 70, 00, 100, 10,145

3.5.10 RADIATION MONITORING INSTRUMENTATION APPLICABILITY:

As shown in Table 3.5.10-1.

OBJECTIVE:

To ensure reliability of the radiation monitoring instrumentation.

SPECIFICATION: The radiation monitoring instrumentation shown in Table 3.5.10-1 shall be OPERABLE with their alarm setpoints within the specified limits.

ACTION:

A. With a radiation monitoring channel alarm setpoint exceeding the value shown in Table 3.5.10-1, adjust the setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.

B. With one or more radiation monitoring channels inoperable, take the ACTION shown in Table 3.5.10-1.

C. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

BASIS:

The OPERABILITY of the radiation monitoring channels ensures that (1) the radiation levels are continually measured in the areas served by the individual channels, and (2) the alarm is initiated when the radiation level trip setpoint is exceeded.

REFERENCES:

(1) NRC letter dated November 1, 1983, from D. G. Eisenhut to to all Pressurized Water Reactors Licensees, NUREG-0737 Technical Specification (Generic Letter No. 83-37).

SAN ONOFRE -

UNIT 1 3.5-28 AMENDMENT NO:

$Z, 120, 145

TABLE 3.5.10-1 RADIATION MONITORING INSTRUMENTATION C)

M MINIMUM CHANNELS APPLICABLE ALARM MEASUREMENT c

INSTRUMENT OPERABLE MODES SETPOINT RANGE ACTION

1.

AREA MONITORS

a. Control Room Area 1

All 1 mR/hr 10 102 mR/hr 25 (R-1231)

b. Spent Fuel Pool Area 1

25 mR/hr 10 102 mR/hr 26 (R-1236)

c.

Containment Radiation 2

1, 2, 3 & 4 10 R/hr 1 - 108 R/hr 27 Monitor-High Range (R-1255, R-1257)

2.

PROCESS MONITORS qn

a. Wide Range Gas Monitor 1

1, 2, 3 & 4 per ODCM 10-'

10s mCi/cc 27 (R-1254)

b. Main Steam Dump and Safety Valve Channels 1/steamline 1, 2, 3 & 4 1 mR/hr (low) 10 104 mR/hr 27 (R-1256A&B, R-1258A&B) 1 R/hr (high) 10- -

104 R/hr

With fuel in the spent fuel pool or building ba

TABLE 3.5.10-1 (Continued)

ACTION STATEMENTS ACTION 25 -

With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement; within Lhour: (1) either initiate and maintain operation of the control room emergency ventilation system in the recirculation mode of operation, or (2) initiate the preplanned alternate method of monitoring and alarming the area radiation.

ACTION 26 -

With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 27 -

With the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirement of Table 3.5.10-1, initiate the preplanned alternate method of monitoring the appropriate parameter(s) within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and:

(1) either restore the inoperable channel(s) to OPERABLE status within 7 days of initiating the preplanned alternate method, or (2) prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2, within 14 days following initiating the preplanned alternate method, outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

SAN ONOFRE -

UNIT 1 3.5-30 AMENDMENT NO: $Z, SI, 1Z0, 145

3.15 [Deleted]

SAN ONOFRE -

UNIT 1 3.15-1 AMENDMENT NO: 70, 110, 145

3.16 RADIOACTIVE GASEOUS EFFLUENTS 3.16.1 (Deleted]

SAN ONOFRE - UNIT 1 3.16-1 AMENDMENT NO:

7$, 130,. 14 5.

3.16.2

[Deleted]

SAN ONOFRE -

UNIT 1 3.16-2 AMENDMENT NO: - #

, di,

~ AI~,145

.3.16.3

[Deleted]

SAN ONOFRE -

UNIT 1 3.16-3 AMENDMENT NO:

70, 00, OZ, 105, 150, 145

3.16.4

[Deleted]

SAN ONOFRE -

UNIT 1 3.16-4 AMENDMENT NO:

70, 00, 0Z, 10$, 130, 145

3.16.5 GAS STORAGE TANK APPLICABILITY:

At all times.

OBJECTIVE:

Limit the amount of radioactivity contained in gas storage tanks.

SPECIFICATION:

A. The quantity of radioactivity contained in each gas storage tank shall be limited to ( 56,000 curies noble gases (considered as Xe-133J.

B. ACTION:

1. With the quantity of radioactive material in any gas storage tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit.

2. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

BASIS:

The tanks included in this specification are those tanks for which the quantity of radioactivity contained is not limited directly or indirectly by another Technical Specification to a quantity that is less than the quantity which provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem in an event of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 mrem.

This is consistent with Branch Technical Position ETSB 11-5 in NUREG 0800, July 1982.

SAN ONOFRE -

UNIT 1 3.16-5 AMENDMENT NO:

7, 0, 110, 145

O 3.16.6 EXPLOSIVE GAS MIXTURE APPLICABILITY:

At all times.

OBJECTIVE:

Limit the amount of explosive gases contained in the gas storage tanks.

SPECIFICATION:

A. The concentration of oxygen in the waste gas holdup system shall be limited to less than or equal to 2%

by volume whenever the hydrogen concentration exceeds 4% by volume.

B.

ACTION:

1. With the concentration of oxygen in the waste gas holdup system greater than 2% by volume but less than or equal to 4% by volume, restore the concentration of oxygen to within the limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

2. With the concentration of oxygen in the waste gas holdup system greater than 4% by volume immediately suspend all additions of waste gases to the system and reduce the concentration of oxygen to less than or equal to 2% by volume without delay.

3. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

BASIS:

This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas holdup system is maintained below the flammability limits of hydrogen and oxygen.

Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

0 SAN ON0FRE -

UNIT 1 3.16-6 AMENDMENT NO:

79. 0, 120,145

3.17

[Deleted]

SAN ONOFRE -

UNIT 1 3.17-1 AMENDMENT NO:

79, 00, SI, 10i, 13a 145

3.18

[Deleted]

SAN ONOFRE -

UNIT 1 3.18-1 AMENDMENT NO:

70, 97, 10B, 130, 145

3.19

[Deleted]

SAN ONOFRE - UNIT 1 3.19-1 AMENDMENT NO:

70, 00, 10, 130, 145

4.1.2

[Deleted]

SAN ONOFRE -

UNIT 1 4.1-13 AMENDMENT NO: 70, 130, 145

4.1.3

[Deleted]

SAN ONOFRE -

UNIT 1 4.1-14 AMENDMENT NO: 7%, 110, 145

4.1.4 CONTAINMENT ISOLATION INSTRUMENTATION APPLICABILITY Applies to instrumentation which actuates the containment sphere isolation valves, containment sphere purge and exhaust valves, and containment sphere instrumentation vent.header valves.

OBJECTIVE:

To ensure reliability of the containment sphere isolation provisions.

SPECIFICATION:

A. Each instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION, and CHANNEL TEST operations for the MODES and at the frequencies shown in Table 4.1.4-1.

B. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation.

BASIS:

The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standard.

The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.

REFERENCES:

(1) NRC letter dated July 2, 1980, from D. G. Eisenhut to all pressurized water reactor licensees.

SAN ONOFRE -

UNIT 1 4.1-15 AMENDMENT NO: $g. U$, 145

TABLE 4.1.4-1 0

CONTAINMENT ISOLATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS ri-i MODES IN WHICH CHANNEL CHANNEL CHANNEL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED Containment Isolation (Valves listed in Table 3.6.2-1) a) Manual N.A.

N.A.

M(1) 1, 2, 3, 4 b) Containment Pressure-High N.A.

R M(2) 1, 2, 3 c) Sequencer Subchannels N.A.

N.A.

M 1, 2, 3, 4 d) Safety Injection

1) Containment Pressure-High N.A.

R M(2) 1, 2, 3

2) Pressurizer Pressure-Low N.A.

R M

1, 2, 3, 4 Purge and Exhaust Isolation (POV-9, POV-10, CV-10, CV-40, CV-116) a) Manual N.A.

N.A.

M(1) 1, 2, 3, 4 i

b) Containment Radioactivity-High S

R M

1, 2, 3, 4 r M

TABLE 4.1.4-1 (Continued)

TABLE NOTATION (1) Manual actuation switches shall be tested at least once per 18 months during shutdown. All other circuitry associated-with manual safeguards actuation shall receive a CHANNEL TEST at least once per 31 days.

(2) The CHANNEL TEST shall include exercising the transmitter by applying either a vacuum or pressure to the appropriate side of the transmitter.

SAN ONOFRE - UNIT 1 4.1-17 AMENDMENT NO: f$, X10. 145

O 4.1.5 ACCIDENT MONITORING INSTRUMENTATION APPLICABILITY:

MODES 1, 2 and 3.

OBJECTIVES:

To ensure the reliability of the accident monitoring instrumentation shown in Table 4.1.5-1.

SPECIFICATION:

Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.1.5-1.

BASIS:

The surveillance requirements specified for these systems ensure that the overall functional capability is maintained comparable to the original design standards.

The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.

References:

(1) NRC letter dated July 2, 1980, from D. G. Eisenhut to all pressurized water reactor licensees.

SAN ONOFRE -

UNIT 1 4.1-18 AMENDMENT NO: SE.

. 130, 145

TABLE 4.1.5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION Pressurizer Water Level M

R Auxiliary Feedwater Flow Indication*

M R

o Auxiliary Feedwater Flow Rate M

R o Steam Generator Water Level (Wide Range)

M R

o Reactor Coolant System Loop Delta-T Indication M

R Reactor Coolant System Subcooling Margin Monitor M

R PORV Position Indicator M

R PORV Block Valve Position Indicator M

R Safety Valve Position Indicator M

R Containment Pressure (Wide Range)

M R

Refueling Water Storage Tank Water Level M

R Containment Sump Water Level (Narrow Range)

M R

Containment Water Level (Wide Range)

M R

Neutron Flux (Wide Range)

M R**

See footnote of Table 3.5.6-1.

- Neutron detectors may be excluded from CHANNEL CALIBRATION.

SAN ONOFRE -

UNIT 1 4.1-19 AMENDMENT NO: 58, 85, 17 If9. 120.

145

4.1.6 PRESSURIZER RELIEF VALVES APPLICABILITY:

Applies to the power operated relief valves (PORVs) and their associated block valves for MODES 1, 2, and 3.

OBJECTIVE To ensure the reliability of the PORVs and block valves.

SPECIFICATION:

A. Each PORV shall be demonstrated OPERABLE:

1. At least once per 31 days by performance of a CHANNEL TEST, which may include valve operation, and

2. At least once per 18 months by performance of a CHANNEL CALIBRATION.

B. Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel, unless the block valve is being maintained closed in order to meet the requirements of Specification 3.1.5.A.

C. The backup nitrogen supply for the PORVs and block valves shall be demonstrated OPERABLE at least once per 18 months by transferring motive power from the normal air supply to the nitrogen supply and operating the valves through a complete cycle of full travel.

BASIS:

The power operated relief valves (PORVs) operate to relieve RCS pressure below the getting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The air supply for both the relief valves and the block valves is capable of being supplied from a backup passive nitrogen source to ensure the ability to seal this possible RCS leakage path.

REFERENCES:

(1) NRC letter dated July 2, 1980, from D. G. Eisenhut to all pressurized water reactor licensees.

SAN ONOFRE -

UNIT 1 4.1-20 AMENDMENT NO: 58, 10, ZZO, 145

4.1.7 PRESSURIZER APPLICABILITY:

Applies to pressurizer heaters and pressurizer water level for MODES 1, 2, and 3.

OBJECTIVE:

To ensure proper pressurizer water volume and to ensure the capability to energize the pressurizer heaters from the emergency diesel generator.

SPECIFICATION:

A. The pressurizer water level shall be determined to be between 5% and 70% at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

B. The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by transferring power from the normal supply to the emergency diesel generator and energizing the heaters.

BASIS:

The requirement that the pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency diesel generator provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation at HOT STANDBY.

REFERENCES:

(1) NRC Letter dated July 2, 1980, from D. G. Eisenhut to all pressurized water reactor licensees.

SAN ONOFRE -

UNIT 1 4.1-21 AMENDMENT NO: 'F$,

Z2 145

4.1.8 AUXILIARY FEEDWATER INSTRUMENTATION APPLICABILITY:

Applies to the auxiliary feedwater instrumentation and interlocks in MODES 1, 2, and 3.

OBJECTIVE:

To ensure reliability of automatic initiation of the auxiliary feedwater system.

SPECIFICATION:

A. Each instrumentation channel shall be demonstrated OPERABLE by the performance of the surveillance requirements specified in Table 4.1.8-1..

BASIS:

The surveillance requirements specified for this instrumentation ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.

REFERENCES:

(1) NRC letter dated July 2, 1980, from D. G. Eisenhut to all pressurized water reactor licensees.

SAN ONOFRE - UNIT 1 4.1-22 AMENDMENT NO: 58, Si, 125 1%0, 145

TABLE 4.1.8-1 C)

AUXILIARY FEEDWATER INSTRUMENTATION rn SURVEILLANCE REQUIREMENTS C__

TRIP ACTUATING DEVICE MODES IN WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST REQUIRED

a. Manual N/A N/A N/A R

1, 2, 3,

b. Automatic Actuation Logic N/A N/A M

N/A 1, 2, 3

c. Steam Generator Water S

R M

N/A 1,2,3 Level-Low

d.

AFW Train Interlocks N/A R

M N/A 1, 2, 3 IT 0n mn

. 4.1.9 AUXILIARY FEEDWATER SYSTEM SURVEILLANCE APPLICABILITY:

Applies to the auxiliary feedwater pumps and valves for MODES 1, 2, and 3.

OBJECTIVE:

To ensure the reliability of the auxiliary feedwater system.

SPECIFICATION:

A. Each auxiliary feedwater pump shall be demonstrated OPERABLE by testing each pump in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR 50.55a(g)(6)(i).

B. At least once per 31 days an inspection shall be made to verify that each non-automatic valve in the emergency flow path is not locked, sealed, or otherwise secured in position is in its correct position.

C. Each auxiliary feedwater Train shall be demonstrated OPERABLE at least once per 18 months by:

1. Verifying that the AFW Train B pump starts as designed automatically upon receipt of an auxiliary feedwater actuation test signal.

2. Verifying that AFW Train A motor driven pump starts as designed automatically upon receipt of auxiliary feedwater actuation AND Train B low flow test signals. Subsequently, verify the pump stops upon receipt of a Train B positive flow test signal.

3. Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after entering MODE 3, verifying that the AFW Train A steam driven pump enters warm-up mode upon receipt of an auxiliary feedwater actuation test signal.

Subsequently, verify pump starts upon receipt of a Train B low flow test signal, and returns to warm-up mode upon receipt of Train B positive flow test signal.

4. Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of actuation test signals.

D. When the reactor coolant system pressure remains less than 500 psig for a period longer than thirty (30) days, flow tests shall be performed to verify the emergency flow paths from the auxiliary feedwater storage tank to each steam generator, using each motor driven auxiliary feedwater pump prior to increasing reactor coolant system pressure above SAN ONOFRE -

UNIT 1 4.1-24 AMENDMENT NO:

OZ, 82, 12 130, 145

500 psig. The flow tests shall be conducted with the auxiliary feedwater system valves in their emergency alignment. Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after entering MODE 3, the steam driven auxiliary feedwater pump shall be similarly tested.

E. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 for the steam driven auxiliary feedwater pump. However, the steam driven AFW pump must be OPERABLE in all other respects.

BASIS:

The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 350aF from normal operating conditions in the event of a total loss of offsite power.

The design of the auxiliary feedwater system further ensures sufficient AFW flow into the intact feedwater lines without exceeding pump run-out or water hammer limits for any applicable design basis event with or without concurrent loss of offsite power and a single active failure.( 2,3)

Specification 4.1.9.A demonstrates the operability of the AFW pumps by testing requirements of the Sec tion XI IST program.(4 ) During normal power operation, pump G-10W is substantial flow tested and pump G-10 and G-10S are tested on minimum flow. During this test differential pressure, pump speeds, and bearing vibration are measured for all 3 pumps. Pump dis charge flow is measured for AFW pump G-10W. At each return from a MODE 3, 4, or 5 shutdown following power operations, pumps G-10 and G-10S are full flow tested unless tested in the previous 92 days. G-10S is tested prior to MODE 3. G-10 is tested in MODE 1 at about 20% power.

REFERENCES:

(1) NRC letter dated July 2, 1980, from D. G. Eisenhut to all pressurized water reactor licensees.

(2) SCE letter dated November 6, 1987, from M. 0.

Medford to NRC Document Control Desk.

(3) SCE letter dated November 20, 1987, from M. 0.

Medford to NRC Document Control Desk.

(4) SCE letter dated January 3, 1991, from F. R.

to NRC Document Control Desk.

SAN ONOFRE -

UNIT 1 4.1-25 AMENDMENT NO: 0, 75, ZZ0. 145

. 4.1.10 AUXILIARY FEEDWATER STORAGE TANK SURVEILLANCE APPLICABILITY:

Applies to the auxiliary feedwater storage tank for MODES 1, 2, and 3.

OBJECTIVE:

To ensure the availability of an adequate auxiliary feedwater supply.

SPECIFICATION:

A.

The auxiliary feedwater storage tank shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying the contained water volume is-within its limits when the tank is the supply source for the auxiliary feedwater pumps.

BASIS:

See Basis for 3.4.4.

SAN ONOFRE -

UNIT 1 4.1-26 AMENDMENT NO:

02, 10, 145

4.1.11 RADIATION MONITORING INSTRUMENTATION APPLICABILITY:

As shown in Table 4.1.11-1 OBJECTIVE:

To ensure the reliability of the radiation monitoring instrumentation.

SPECIFICATION:

A. Each radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL TEST operations for the MODES and at the frequencies shown in Table 4.1.11-1.

BASIS:

The surveillance requirements specified for these systems ensure that the overall functional capability is maintained comparable to the original design standards.

The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.

REFERENCES:

(1) NRC letter dated November 1, 1983, from D. G. Eisenhut to all Pressurized Water Reactor Licensees, NUREG-0737 Technical Specification (Generic Letter No. 83-37).

SAN ONOFRE - UNIT 1 4.1-27 AMENDMENT NO:

S2. 10 145

TABLE 4. 1.

11-1 0

Co RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS m

CHANNEL CHANNEL CHANNEL APPLICABLE INSTRUMENT CHECK CALIBRATION TEST MODES

1.

AREA MONITORS

a. Spent Fuel Pool Area S

R M

(R-1236)

b. Control Room Area S

R M

All (R-1231)

c. Containment Radiation S

R M

1, 2, 3 & 4 Monitor-High Range (R-1255, R-1257)

2. PROCESS MONITORS

a. Wide Range Gas Monitor 1, 2, 3 & 4 (R-1254)

b. Main Steam Dump and Safety S

R M

1, 2, 3 & 4 Valve Channels (R-1256A&B, R-1258A&B)

M X

bA

See footnote of Table 3.5.10-1

- In accordance with Table 4.1.3.1 surveillance requirements for this instrument channel.

4.1.12 REACTOR COOLANT SYSTEM VENTS APPLICABILITY:

MODES 1, 2, 3 and 4.

OBJECTIVE:

To ensure the reliability of the reactor coolant vent system.

SPECIFICATION:

Each reactor coolant system vent path shall be demonstrated OPERABLE at least once per 18 months by:

1. Verifying all manual isolation valves in each vent path are locked in the open position.

2. Cycling each valve in the vent path through at least one complete cycle of full travel from the control room during COLD SHUTDOWN or REFUELING.

3. Verifying flow through the reactor coolant vent system vent paths during venting during COLD SHUTDOWN.

BASIS:

See basis for 3.1.7, Reactor Coolant System Vents.

REFERENCES:

NRC letter dated November 1, 1983, from D. G. Eisenhut to all Pressurized Water Reactor Licensees, NUREG-0737 Technical Specifications (Generic Letter No. 83-37).

SAN ONOFRE -

UNIT 1 4.1-29 AMENDMENT NO: $, 30. 145

4.1.13 LEAKAGE AND LEAKAGE DETECTION SYSTEMS APPLICABILITY:

Applies to the reactor coolant leakage and detection systems delineated in Specification 3.1.4.

OBJECTIVE:

To ensure the reactor coolant system leakage limits are maintained and to ensure the OPERABILITY of those systems that are used to detect leakage from the reactor coolant system.

SPECIFICATION:

A. Reactor Coolant System leakage shall be demonstrated to be within limits by:

1. Monitoring the containment atmosphere radioactivity at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

3. Monitoring the steam generator blowdown effluent radioactivity at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4. Monitoring the containment sump level indicator (LIS 2001 or 3001) at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

B. The leakage detection systems shall be demonstrated OPERABLE by the performance of CHANNEL CHECK, SOURCE CHECK, CHANNEL TEST, AND CHANNEL CALIBRATION at the frequencies specified in Table 4.1.13-1; BASIS:

The monitoring of reactor coolant system leakage and maintenance of OPERABILITY of the reactor coolant leakage detection systems will assure that the sources of leakage are monitored and/or identified. The methods described above provide an acceptable means of verifying the OPERABILITY required by Specification 3.1.4.

REFERENCES:

1. SEP Topic V-5, Reactor Coolant Pressure Boundary Leakage, NUREG-0829, December 1986

2. Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973

3. Standard Technical Specifications for Westinghouse Pressurized Water Reactors, Revision 4, NUREG-0452 SAN ONOFRE -

UNIT 1 4.1-30 AMENDMENT NO:

ZZ0, Z30, 145

TABLE 4.1.13-1 LEAKAGE DETECTION SYSTEMS CHANNEL SOURCE CHANNEL CHANNEL INSTRUMENT CHECK CHECK TEST CALIBRATION

1. Containment Atmosphere D

M N/A R

Particulate Monitor (R1211)

2. Containment Atmosphere Gaseous Monitor (R1212)

3. Sphere Sump Level N/A N/A N/A R

Control System (LS80 and 82)

4. Containment Sphere Sump Level Monitor (LIS 2001 N/A N/A and 3001)

5.

Steam Generator Blowdown Effluent Monitor (R1216)

In accordance with Table 4.1.3.1, surveillance requirements for this instrument channel.

- In accordance with Table 4.1.5-1, surveillance requirements for these instrument channels.

- - In accordance with Table 4.1.2.1, surveillance requirements for this instrument channel.

. 4.5 1Deleted]

SAN ON0FRE -

UNIT 1 4.5-1 AMENDMENT NO: 1S, 70, 120,145

4.6 RADIOACTIVE GASEOUS EFFLUENTS 4.6.1

[Deleted]

SAN ONOFRE -

UNIT 1 4.6-1 AMENDMENT NO: 38, 7S, 130 145

4.6.2

[Deleted]

SAN ONOFRE -

UNIT 1 4.6-2 AMENDMENT NO:

7S, 79, 110, 145

4.6.3

[Deleted]

SAN ONOFRE - UNIT 1 4.6-3 AMENDMENT NO: X9, 70, 110 145

4.6.4

[Deleted]

SAN ONOFRE -

UNIT 1 4.6-4 AMENDMENT NO: 75, 130, 145

4.6.5 GAS STORAGE TANK APPLICABILITY:

At all times.

OBJECTIVE:

To verify the quantity of radioactive material contained within the gas storage tanks.

SPECIFICATION:

The quantity of radioactivity material contained in each gas storage tank shall be determined to be within the limit specified in Specification 3.16.5 at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when radioactive materials are being added to the tank.

BASIS:

Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem.

This is consistent with Branch Technical Position ETSB 11-5 in NUREG-0800, July 1981.

SAN ONOFRE -

UNIT 1 4.6-5 AMENDMENT NO: 79, 170, 145

4.6.6 EXPLOSIVE GAS MIXTURE APPLICABILITY:

At all times.

OBJECTIVE:

Limit the amount of explosive gases contained in the gas storage tanks.

SPECIFICATION:

The concentrations of hydrogen and/or oxygen in the waste gas holdup system shall be determined to be within the limits specified in Specification 3.16.6 by analyzing grab samples of the waste gas holdup system contents at the waste gas decay tank in service daily and every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during degassing. Degassing is defined as the process to reduce reactor coolant system (RCS) dissolved H gas concentration in preparation for refueling or for Opening the reactor coolant system.

BASIS:

This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas holdup system is maintained below the flammability limits of hydrogen and oxygen.

Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the release of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

SAN ONOFRE - UNIT 1 4.6-6 AMENDMENT NO: 7, Igg, 145

. 4.17

[Deleted]

SAN ONOFRE -

UNIT 1 4.17-1 AMENDMENT NO:

70, 120, 145

4.18

[Deleted]

SAN ONOFRE -

UNIT 1 4.18-1 AMENDMENT NO:

7, 120, 145

4.19

[Deleted]

SAN ONOFRE -

UNIT 1 4.19-1 AMENDMENT NO:

70, 101, 12 145

6.5.2.5 Proposes tests and experiments which affect station nuclear safety and are not addressed in the FSAR or Technical Specifications shall be reviewed by the Station Manager, or members of the site/station management staff previously designated by the Vice President and Site Manager, Nuclear Generation Site. Documentation of these activities.shall be provided to the Vice President and Site Manager, Nuclear Generation Site, and the NSG.

6.5.2.6 Recommended changes to the station security plan shall be approved by the Station Manager and transmitted to the Vice President and Site Manager, Nuclear Generation Site, and to the NSG; implementing procedures shall be prepared and approved in accordance with Specification 6.8.

6.5.2.7 Recommended changes to the station emergency plan shall be approved by the Station Manager and transmitted to the Vice President and Site Manager, Nuclear-Generation Site, and to the NSG; implementing procedures shall be prepared and approved in accordance with Specification 6.8.

6.5.2.8 The Station Manager shall assure the performance of a review by a qualified individual/organization of every uncontrolled or unplanned release of radioactivity to the environs including the preparation and forwarding of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence to the Vice President and Site Manager, Nuclear Generation Site, to the NSG, and to the Station Manager.

6.5.2.9 The Station Manager shall assure the performance of a review by a qualified individual/organization and may designate the approval of changes to the PROCESS CONTROL PROGRAM, OFFSITE DOSE CALCULATION MANUAL, and radwaste treatment systems.

Documentation of these activities shall be provided to the Vice President and Site Manager, Nuclear Generation Site, to the NSG, and to the Station Manager.

6.5.2.10 Documentation of each of the activities performed under Specifications 6.5.2.1 through 6.5.2.9 shall be maintained.

SAN ON0FRE 1 UNIT 1 6.5-4 AMENDMENT NO: 50, 00, 09, 7, SI, ZZg, 145

(ii)

Identification of the procedures used to measure the values of the critical parameters, (iii)

Identification of process sampling points, (iv)

Procedures for the recording and management of data, (v)

Procedures defining corrective actions for off control point chemistry conditions, and (iv)

A procedure identifying (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiated corrective action.

e. Post-Accident Sampling A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. The program shall include:

(i)

Training of personnel, (i)

Procedures for sampling and analysis, and (iii)

Provisions for maintenance of sampling and analysis equipment.

f. Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

(1) Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM; (2) Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to 10 CFR Part 20, Appendix B, Table II, Column 2;

(3) Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the ODCM; SAN ONOFRE -

UNIT 1 6.8-3 AMENDMENT NO:

01, 110 145

(4) Limitations on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conforming to Appendix I to 10.CFR Part 50; (5) Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year, respectively, in accordance with the methodology and parameters in the 00CM at least every 31 days; (6) Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50; (7) Limitations of the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the SITE BOUNDARY conforming to the doses associated with 10 CFR Part 20, Appendix 8, Table II, Column 1; (8) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50; (9) Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radioactive nuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50; (10) Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.

g. Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be contained in the 00CM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

SAN ONOFRE -

UNIT 1 6.8-4 AMENDMENT NO:145

g. Radiological Environmental Monitorint Program (Cont.)

(1) Monitoring, sampling, analysis, and reporting of radiation and radionuclices in the environment in accordance with the methodology and parameters in the ODCM.

(2) A Land Uses Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of this census, and (3) Participation in an Interlaboratory Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

SAN ONOFRE -

UNIT 1 6.8-5 AMENDMENT NO: 145

inspection, routine maintenance, special maintenance (describe main tenance), waste processing, and refueling). The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20 percent of the individual total dose need not be accounted for.

In the aggregate, at least 80 percent of the total whole body dose received from external sources shall be assigned to specific major work functions.

Reports required on an annual basis shall include the results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.1.1. The following information shall be included in these reports: (1) Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (2)

Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis whle limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than limit. Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) Graph of the 1-131 concentration and one other radioiodine isotope concentration in microcuries per gram as a func tion of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activ ity of the primary coolant exceeded the radioiodine limit.

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT*

6.9.1.6 The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year. The report shall include sum maries, interpretation, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period.

The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appen dix I to 10 CFR Part 50.

6.9.1.7

[Deleted]

A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

SAN ONOFRE - UNIT 1 6.9-2 AMENDMENT NO: 12, IS, Z0, 79, SI, 95, ZZ0, 145

SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT*

6.9.1.8 Routine radioactive effluent release reports covering the operation of the unit during the previous 6 months of operation shall be submitted withn 60 days after January 1 and July 1 of each year.

6.9.1.9

[Deleted]

MONTHLY OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to pressurizer safety and relief valves, shall be submitted to the Nuclear Regulatory Commission, on a monthly basis, no later than the 15th of each month following the calendar month covered by the report.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Nuclear Regulatory Commission within the time period specified for each report.

A single submittal may be made for multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases radioactive material from each unit.

[Deleted]

SAN ONOFRE - UNIT 1 6.9-3 AMENDMENT NO: 12, II 0,

71, 01, 90, 105, 1$0, 145

f. Records of reactor tests and experiments.

g. Records of training and qualification for current members of the unit staff.

h. Records of in-service inspections performed pursuant to these Technical Specifications.

i. Records of Quality Assurance activities not included in 6.10.1 that are required by the QA Manual.

j. Records of reviews performed for changes made to procedures or equipment or reviews or tests and experiments pursuant to 10 CFR 50.59.

k. Records of OSRC meetings and NSG reports.

1. Records of the service lives of all safety related hydraulic and mechanical snubbers including the date at which the service life commences and associated installation and maintenance records.

m. Records of secondary water sampling and water quality.

n. Records of review performed for changes made to the OFFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM.

SAN ONOFRE -

UNIT 1 6.10-2 AMENDMENT NO:

Z2, OZ, O1, SI. 130, 145

6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 The PCP shall be approved by the Commission prior to implemtation.

6.13.2 Licensee-initiated changes to the PCP:

1. Shall be documented and records of reviews performed shall be retained as required by specification 6.10.2.

a. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s);

b. A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations; and C.

Documentation of the fact that the change has been reviewed and found acceptable pursuant to 6.5.2.

.2. Shall become effective upon review and approval pursuant to 6.5.2. by the Station Manager.

SAN ONOFRE -

UNIT 1 6.13-1 AMENDMENT NO: 58, 70. OI, V0, 145

6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6.14.1 The ODCM shall be approved by the Commission prior to implementation.

6.14.2 Licensee-initiated changes to the ODCM:

1. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change(s) was made effective. This submittal shall contain:

a. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s);

b. A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations; and

c. Documentation of the fact that the change has been reviewed and found acceptable pursuant to 6.5.2.

2. Shall become effective upon review and approval pursuant to 6.5.2. by the Station Manager.

3. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as part of or concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made.

Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g.

month/year) the change was implemented.

SAN ONOFRE - UNIT 1 6.14-1 AMENDMENT NO:

58, 79, SI, 10, 145

6.15 [Deleted]

SAN ONOFRE -

UNIT 1 6.15-1 AMENDMENT NO: 58, 70, 01, 120, 145

ML13309A243

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