Text

Amend 4 to License NPF-10,adding Three Valves Omitted from Table of Motor Operated Valves,Adding Special Test Exceptions,Correcting Errors & Clarifying Administrative Controls & Air Lock Seal Pressure
	ML20062C024
Person / Time
Site:	San Onofre
Issue date:	07/16/1982
From:	Miraglia F; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20062C028	List: ML20062C024; ML20062C042;
References
	NUDOCS 8208050133
	Download: ML20062C024 (104)
	v • d • e

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i SOUTHERN CALIFORNIA EDIS0N COMPANY SAfl DIEGO GAS AND ELECTRIC C0'tAPHY THE CITY OF RIVERSIDE, CALIFORNIA THE CITY OF ANAHEIH, CALIFORNIA DOCKET NO. 50-361 SAN OH0FRE NUCLEAR GENERATING STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendnent No. 4 License No. NPF-10 l

1.

The Nuclear Regulatory Comission (the Connission) has found that:

A.

The application for anendment to the San Onofre Nuclear Generating Station, Unit 2 (the facility) Facility Operating License No. NPF-10 filed by the Southern California Edison Company on behalf of itself and San Diego Gas and Electric Company, The City of Riverside and The City of Anahein, California (licensees) dated May 14, 1982, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulation as 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 regulations of the Comission; 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 Comission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this license amendment will not be inimical to the comon defense and security or to the health and safety of the public; E.

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

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2.

Accordingly, the license is amended by changes to the Technical Specific-ations as indicated in the attachment to this license amndment, and paragraph 2.C(2) of Facility Operating License No. flPF-10 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environ-r.cntal Protection Plan contained in Appendix B, as revised through Amendnent No. 4, are hereby incorporated in the license.

SCE shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY C0t311SS10tl

' Original signed by Frank J. Miraglia Frank J. Miraglia, Chief Licensing Branch No. 3 Division of Licensing

Attachment:

Changes to the Technical Specificatinns Date of Issuance: Jul 161332 1

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WASHINGTON, D. C. 20555 SOUTHERN CALIFORNIA EDIS0N COMPANY SAN DIEGO GAS AND ELECTRIC COMAPNY THE CITY OF RIVERSIDE, CALIFORNIA THE CITY OF ANAHEIM, CALIFORNIA DOCKET NO. 50-361 SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 4 License No. NPF-10 1.

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

A.

The application for amendment to the San Onofre Nuclear Generating Station, Unit 2 (the facility) Facility Operating License No. NPF-10 filed by the Southern California Edison Company on behalf ~ of itself and San Diego Gas and Electric Company, The City of Riverside and The City of Anaheim, California (licensees) dated May 14, 1982, complies with the standards and requircaents of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulation as 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 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 set forth in 10 CFR Chapter I; D.

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

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

i 4

2.

Accordingly, the license is amended by changes to the Technical Specific-ations as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Operating License No. NPF-10 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environ-mental Protection Plan contained in Appendix B, as revised through Amendment No. 4, are hereby incorporated in the license.

SCE shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Frank J. Miraglia, Chief Licensing Branch No. 3 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance:M 161982

a AMENDMENT TO LICENSE AMENDMENT NO. 4 FACILITY OPERATING LICENSE NO. NPF-10 DOCKET N0. 50-361 Replace the following pages of the Appendix A Technical Specification with enclosed pages.

The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

The corresponding overleaf pages are also provided to maintain document completeness.

Overleaf Amended P_aSe Page VI V

XVII XVIII 3/4 1-11 3/4 1-12 3/4 1-13 3/4 1-14 3/4 1-19 3/4 1-20 3/4 1-25 3/4 3-3 3/4 3-4 3/4 3-12 3/4 3-11 3/4 3-13 3/4 3-14 3/4 3-16 3/4 3-15 3/4 3-20 3/4 3-19 3/4 3-21 3/4 3-22 3/4 3-51 3/4 3-52 3/4 3-53 3/4 3-53a 3/4 4-2 3/4 4-1 3/4 4-4 3/4 4-3 3/4 4-6 3/4 4-5 3/4 4-7 3/4 4-8 3/4 4-31 3/4 4-32 3/4 4-34 3/4 4-33 3/4 6-5 3/4 6-6 3/4 6-9 3/4 6-10 3/4 6-12 3/4 6-11 3/4 6-12a 3/4 6-24 3/4 6-23 3/4 7-3 3/4 7-4 3/4 7-10 3/4 7-9 3/4 8-31 3/4 8-32 3/4 8-33 3/4 8-34 3/4 8-35 3/4 10-4 3/4 10-3 3/4 10-5 3/4 10-6 3/4 10-8 3/4 12-5 3/4 12-6 3/4 10-7

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REPLACEMENT PAGES (CONTINUED)

AMENDMENT NO. 4 Overl ea f Amended Page Page B3/4 7-7 B3/4 8-1 B3/4 8-2 6-2 6-1 6-3 6-4 6-4a 6-5 6-6 6-7 6-8 6-9 6-10 6-13 6-14 6-23 6-24 7-1 7-2 B3/4 10-2 B3/4 10-1

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I INDEX LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS

{

SECTION PAGE HOT SHUTD0WN............................................

3/4 4-3 COLD SHUTDOWN - Loops Filled............................

3/4 4-5 COLD SHUTDOWN - Loops Not Filled........................

3/4 4-6 t

s 3e 1 4.2 SAFETY VALVES - 0PERATING...............................

3/4 4-7 i

i 3/4.4.3 PRESSURIZER.............................................

3/4 4-8 l

3/4.4.4 STEAM GENERATORS..................................'......

3/4 4-9 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS............................

3/4 4-16 r

OPERATIONAL LEAKAGE...................

3/4 4-17 3/4.4.6 CHEMISTRY.............................................

3/4 4-20 l

3/4.4.7 SPECIFIC ACTIVITY...............'.4...'......g

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3/4 4-23 4

3/4.4.8 PRESSURE / TEMPERATURE LIMITS

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REACTOR COOLANT SYSTEM......................

3/4 4 '

PRESSURIZER - HEATUP/C00LDOWN......................l..

3/4 4-31 l: }

OVERPRESSURE PROTECTION SYSTEMS s

s i N, RCS TEMPERATURE 5 235 F............

3/4 4 ;

l RCS TEMPERATURE >;235 F..........,g 3/4 4-33 i

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3/4.4.9 STRUCTURAL INTEGRITY....................................

3/4 4-34.

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3/4.5 EMERGENCY CORE COOLING SYSTFMS 7,

3/4.5.1 SAFETY INJECTION TANKS..............~...

3/4 S-1 i

3/4.5.2 ECCS SUBSYSTEMS - Tavg 350 F..........................

3/4 5-3 s -

3 3/4.5.3 ECCS SUBSYSTEMS - Tavg 1< 350 F...

3/4 5-7.

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N REFUELING WATER STORAG'E TANA........

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY................................

3/4 6-1 CONTAINMENT LEAKAGE..................................

3/4 6-2 s

CONTAINMENT AIR L0CKS................................

3/4 6-5 INTERNAL PRESSURE....................................

3/4 6-7 AIR TEMPERATURE......................................

3/4 6-8 CONTAINMENT STRUCTURAL INTEGRITY.....................

3/4 6-9 CONTAINMENT VENTI LATION SYSTEM.......................

3/4 6-13 3/4.6 2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM.............................

3/4 6-14 IODINE REMOVAL SYSTEM................................

3/4 6-16

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CONTAINMENT COOLING SYSTEM.........................

3/4 6-17 3/4.6.3 CONTAINMENT ISOLATION VALVES............................

3/4 6-18 3/4.6.4 COMBUSTIBLE GAS CONTROL HYDROGEN MONITORS...................................

3/4 6-26 ELECTRIC HYDROGEN REC 0MBINERS........................

3/4 6-27 CONTAINMENT DOME AIR CIRCULAT0RS.....................

3/4 6-28

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s INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY...............................................

6-1 6.2 ORGANIZATION 6.2.1 0FFSITE.................................................

6-1

- 6.2.2 UNIT STAFF.,............................................

6-1 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP....................

6-5 6.2.4 SHIFT TECHNICAL ADVIS0R.................................

6-5 6.3 UNIT STAFF QUALIFICATIONS....................................

6-5 6.4 TRAINING.....................................................

6-6

6. 5 REVIEW AND AUDIT 6.5.1 ONSITE REVIEW COMMITTEE FUNCTION.............................................

6-6 COMPOSITION..........................................

6-6 3

A LT E RNAT E S...........................................

6-6 MEETING FREQUENCY....................................

6-7 QU0RUM...............................................

6-7 RESP 0NSIBILITIES.....................................

6-7 AUTHORITY.......................................

6-8 REC 0RDS..............................................

6-8

' 6.5.2 TECHNICAL REVIEW AND CONTR0L............................

6-8 6.5.3 NUCLEAR SAFETY GROUP FUNCTION................................................

6-9 COMPOSITION.............................................

6-10 CONSULTANTS.............................................

6-10 REVIEW..................................................

~6-10 AUDITS..................................................

6-11 SAN ONOFRE-UNIT 2 XVII

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INDEX i

ADMINISTRATIVE CONTROLS SECTION PAGE AUTHORITY...............................................

6-12 REC 0RDS.................................................

6-12 6.6 REPORTABLE OCCURRENCE ACTI0N.................................

6-13 3

i 6.7 SAFETY LIMIT VIOLATION.......................................

6-13 J

4 6.8 PROCEDURES AND PR0 GRAMS.................

6-13 1

6.9 REPORTING REQUIREMENTS r

6.9.1 ROUTINE AND REPORTABLE OCCURRENCES......................

6-15 4

STARTUP REP 0RT.......................................

6-16 ANNUAL REP 0RTS.......................................

6-16 f

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT...

6-17 3

SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT.......

6-17 i

MONTHLY OPEF.ATING REP 0RT.............................

6-19 REPORTABLE OCCURRENCES...............................

6-19 PROMPT NOTIFICATION WITH WRITTEN FOLLOWUP............

6-19 THIRTY DAY WRITTEN REP 0RTS...........................

6-21 1

HAZARDOUS CARGO TRAFFIC REP 0RT.......................

6-21 6.9.2 SPECIAL REP 0RTS.........................................

6-21 2

6.10 RECORO RETENTION...........................................

6-21 6.11 RADIATION PROTECTION PR0 GRAM................................

6-23 4

6.12 HIGH RADIATION AREA.........................................

6-23 6.13 PROCESS CONTROL PROGRAM (PCP)...............................

6-24 6.14 0FFSITE DOSE CALCULATION MANUAL.............................

6-25 6.15 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS........

6-25 t

- 7.1 SPECIAL TEST PR0 GRAM.........................................

7-1 SAN ONOFRE-UNIT 2 XVIII Amendment No. 4

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REACTIVITY CONTROL SYSTEMS BORIC ACID MAKEUP PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 At least the boric acid makeup pump (s) in the boron injection flow path (s) required OPERABLE pursuant to Specification 3.1.2.2a shall be OPERABLE and capable of being powered from an OPERABLE emergency bus if the flow path through the boric acid pump (s) in Specification 3.1.2.2a is OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one boric acid makeup pump required for the boron injection flow path (s) pursuant to Specification 3.1.2.2a inoperable, restore the boric acid makeup pump to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN equivalent to at least 2% delta k/k at 200 F; restore the above required boric acid makeup pump (s) to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE0VIREMENTS 1

4.1.2.6 No additional Surveillance Requirements other than those required by Specification 4.0.5.

SAN ONOFRE-UNIT 2 3/4 1-11

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REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCE - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.7 As a minimum, one of the following borated water sources shall be OPERABLE:

One boric acid makeup tank and at least one associated heat tracing a.

circuit with the tank contents in accordance with Figure 3.1-1.

b.

The refueling water storage tanks with:

1.

A minimum borated water volume of 5465 gallons above the ECCS suction connection, 2.

A minimum boron concentration of 1720 ppm, and 3.

A solution temperature between 40 F and 100*F.

1 APPLICABILITY: MODES 5 and 6.

J ACTION:

With no borated water sources OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity change 3.

a SURVEILLANCE REQUIREMENTS t

4.1.2.7 The above required borated water source shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1.

Verifying the boron concentration of the water, 2.

Verifying the contained borated water volume of the tank, and 3.

Verifying the boric acid makeup tank solution temperature when it is the source of borated water.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when it is the source of borated water when the outside air temperature is less than 40 F or greater than 100 F.

l SAN ON0FRE-UNIT 2 3/4 1-12 Amendment No. 4

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Figure 3.1 1 MINIMUM BORIC ACID STORAGE TANK VOLUME AND TEMPERATURE AS A FUNCTION OF STORED BORIC ACID CONCENTRATION SAN ONOFRE-UNIT 2 3/4 1-13

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REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR 0PERATION

3. L 2. 8 Each of the following borated water sources shall be OPERABLE:

a.

At least one boric acid makeup tank and at least one associated heat tracing circuit with the contents of the tanks in accordance with Figare 3.1-1,'and b.

The refuelis g water storage tank with:

1.

A minimum contained borated water volume of 362,800 gallons above the ECCS suction connection, 2.

Between 1720 and 2300 ppm of boron, and 3.

A solution temperature between 40 F and 100 F.

l APPLICABILITY:

MODES 1, 2, 3 and 4.

ACTION:

With the above required boric acid makeup tank inoperable, restore

. a.

the tank to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHbTDOWN MARGIN equivalent to at least 2% delta k/k at 200 F; restore the above required boric acid makeup tank to OPERABLE status within the next 7 days or be in COLD SHUTOOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With the refueling water tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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SURVEILLANCE REQUIREMENTS 4.1.2.8 Each borated water sources shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1.

Verifying the boron concentration in the water, 2.

Verifying the contained borated water volume of the water source, and 3.

Verifying the boric acid makeup tank solution temperature.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when the outside air temperature is less than 40 F or greater than 100 F.

s SAN ONOFRE-UNIT 2 3/4 1-14 Amendment No. 4 s

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POSITION INDICATOR CHANNEL - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1. 3. 3 At least one CEA Reed Switch Position Transmitter indicator channel shall be OPERABLE for each shutdown, regulating or part length CEA not fully inserted.

APPLICABILITY: MODES 3*, 4* and 5*.

ACTION:

With less than the above required po.sition indicator channel (s) OPERABLE, immediately open the reactor trip breakers.

SURVEILLANCE REQUIREMENTS i

4.1.3.3 Each of the above required CEA Reed Switch Position Transmitter indicator channel (s) shall be determined to be OPERABLE by performance of a CHANNEL FUNCTIONAL TEST at least once per 18 months.

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i With the reactor trip breakers in the closed position.

SAN ONOFRE-UNIT 2 3/4 1-19

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REACTIVITY CONTROL SYSTEMS CEA OROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full length (shutdown and regulating) CEA drop time, from a withdrawn position greater than or equal to 145 inches, shall be less than or equal to 3.0 seconds from when the electrical power is interrupted to the CEA drive mechanism until the CEA reaches its 90 percent insertion position with:

a.

T greater than or equal to 520 F, and 3yg b.

All reactor coolant pumps operating.

APPLICABILITY:

MODES 1 and 2.

ACTION:

With the drop time of any full length CEA determined to exceed the above limit, be in at least HOT STANDBY within six hours.

SURVEILLANCE REQUIREMENTS 4.1.3.4 The CEA drop time of full length CEAs shall be demonstrated through measurement price to reactor criticality:

I a.

For all CEAs following each removal and reinstallation of the reactor vessel head, b.

For specifically affected individuals CEAs following any maintenance on or modification to the CEA drive system which could affect the drop time of those specific CEAs, and c.

At least once per 18 months.

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SAN ON0FRE-UNIT 2 3/4 1-20 Amendment No. 4 L.

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REACTIVITY CONTROL SYSTEMS PART LENGTH CEA INSERTION LIMITS 1

LIMITING CONDITION FOR OPERATION 3.1.3.7 The position of the part length CEA group shall be:

a.

withdrawn to > 145" or; b.

restricted to prevent the neutron absorber section of the part length CEA group from covering the same axial segment (< 145") of the fuel assemblies for a period in excess of 7 EFPD out of any 30 EFPD period.

APPLICABILITY:

MODES 1 and 2.

ACTION:

With the neutron absorber section of the part length CEA group covering any same axial segment of the fuel assemblies for a period exceeding 7 EFPD out of any 30 EFPD period, either:

Reposition the part length CEA group to ensure no neutron absorber a.

section of the part length CEA group is covering the same axial segment of the fuel assemblies within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or b.

Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

i 4

SURVEILLANCE REQUIREMENTS 4.1.3.7 The position of the part length CEA group shall be determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SAN ON0FRE-UNIT 2 3/4 1-25 Amendment No. 4 p

y i-

y,,m 7

,9m-..

-c.

s m.

w

t_

TABLE 3.3-1 s

2 REACTOR PROTECTIVE INSTRUMENTATION o8 MINIMUM 39 TOTAL NO.

CHANNELS CHANNELS APPLICABLE-

{

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE H0 DES ACTION 2

1.

Manual Reactor Trip 2 sets of 2 1 set of 2 2 sets of 2 1, 2 1

]

2 sets of 2 1 set of 2 2 sets of 2 3*, 4*, 5*

7A 2.

Linear Power Level - High 4-2 3

1, 2 2#, 3#

3.

Logarithmic Power Level-High a.

Startup and Operating 4

2(a)(d) 3 1, 2 2#, 3#

4 2

3 3*, 4*, 5*

7A b.

Shutdown 4

0 2

3,4,5 4

i 4.

Pressurizer Pressure - High 4

2 3

1, 2 2#, 3#

l!

5.

Pressurizer Pressure - Low 4

2(b) 3 1, 2 2#, 3#

6.

Containment Pressure - High 4

2 3

1, 2 2#, 3#

7.

Steam Generator Pressure - Low 4/SG 2/SG 3/SG 1, 2 2#, 3#

w 8.

Steam Generator Level - Low 4/SG 2/SG 3/SG 1, 2 2#, 3#

1, 9.

Local Power Density - High 4

2(c)(d) 3 1, 2 2#, 3#

10. DNBR - Low 4

2(c)(d) 3 1, 2 2#, 3#

11. Steam Generator Level - High 4/SG 2/SG 3/SG 1, 2 2#, 3#

12. Reactor Protection System Logic 4

2 3

1, 2 2#, 3#

7 3*, 4*, 5*

7A

13. Reactor Trip Breakers 4

2(f) 4 1, 2 5

4 3*, 4*, 5*

7A

14. Core Protection Calculators 4

2(c)(d) 3 1, 2 2#, 3#, 7

15. CEA Calculators 2

1 2(e) 1, 2 6, 7

16. Reactor Coolant flow - Low 4/SG 2/SG 3/SG 1, 2 2#, 3#

017. Seismic - High 4

2 3

1, 2 2#, 3#

18.

Loss of Load 4

2 3

1(g) 2#, 3#

a TABLE 3.3-1 (Continued)

TABLE NOTATION 2

To be OPERABLE prior to first exceeding 5% RATED THERMAL POWER.

With the protective system trip breakers in the closed position, the CEA drive system capable of CEA withdrawal, and fuel in the reactor vessel.

The provisions of Specification 3.0.4 are not applicable.

~4 (a) Trip may be manually bypassed above 10 % of RATED THERMAL POWER; bypass shall,g%ofRATEDTHERMALPOWER.e automatically removed when THERMAL POWE to 10 (b) Trip may be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is greater than or equal to 400 psia.

~4 (c) Trip may be manually bypassed below 10 % of RATED THERMAL POWER; bypass shallbeaug%ofRATEDTHERMALPOWER.matically removed when THERMAL POWE equal to 10 During testing pursuant to Special Test Exception 3.10.3, trip may be manually bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically' removed when THERMAL POWER is greater than or equal to 5% of RATED THERMAL POWER.

l (d) Trip may be bypassed during testing pursuant to Special Test Excepti.on 3.10.3.

(e) See Special Test Exception 3.10.2.

(f) Each channel shall be comprised of two trip breakers; actual trip logic shall be one-out-of-two taken twice.

t (g) Trip may be bypassed below 55% RATED THERMAL POWER.

ACTION STATEMENTS ACTION 1 With the number of channels OPERABLE one less'than required by the Minimum Channels OPERABLE requirement, restore the inoperable channei to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the protective system trip breakers.

ACTION 2 -

With the number of channels OPERABLE one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may continue provided the inoperable channel is placed in the bypassed or tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

If the inoperable channel is bypassed, the desirability of maintaining this channel in the bypassed condition shall be reviewed in accordance with Specification 6.5.1.6e.

The channel shall be returned to l

OPERABLE status no later than during the next COLD SHUTDOWN.

SAN ONOFRE-UNIT 2 3/4 3-4 Amendment No. 4

L TABLE 4.3-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS E?

-4.

o m

CilANNEL MODES FOR WillCil i

ch CilANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CilECK CALIBRATION TEST IS REQUIRED N

13.

Reactor Trip Breakers N.A.

N.A.

M,(12) 1, 2, 3*,

4*, 5*

i 14.

Core Protection Calculators S

D(2,4),5(7)

M(11),R(6) 1, 2 i

R(4,5,10),M(8) l 15.

CEA Calculators S

R M,R(6) 1, 2 16.

Reactor Coolant flow-Low S

R M

1, 2

@l7.

Seismic-High S

R M

1, 2 R

18.

Loss of Load S

N.A.

M 1 (9)

T O

l 8

i%

n O

a i

2 e

<e

TABLE 4.3-1 (Continued)

TABLE NOTATION To be OPERABLE prior to first exceeding 5% RATED THERMAL POWER.

U O

With reactor trip breakers in the closed position and the CEA drive system capable of CEA withdrawal.

(1)

Each startup or when required with the reactor trip breakers closed and the CEA drive system capable of rod withdrawal, if not performed in the previous 7 days.

(2) -

Heat balance only (CHANNEL FUNCTIONAL TEST not included), above 15%

of RATED THERMAL POWER; adjust the Linear Power Level signals and the CPC addressable constant multipliers to make the CPC delta T power and CPC nuclear power calculations agree with the calorimetric calculat% if ibA:.a d!1 ruta.cc :s g/ca.u ther. i.L Uw ing Frf r$ICS TESTS, these daily calibrations may be suspended provided these cali-brations are performed upon reaching each major test power plateau and prior to proceeding to the next major test power plateau.

(3)

Above 15% of RATED THERMAL POWER, verify that the linear power subchannel gains of the excore detectors are consistent with the values used to estab-lish the shape annealing matrix clements in the Core Protection Calculators.

(4) -

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(5) -

After each fuel loading and prior to exceeding 70% of RATED THERMAL POWER, the incore detectors shall be used to determine the shape annealing matrix elements and the Core Protection Calculators shall use these elements.

(6)

This CHANNEL FUNCTIONAL TEST shall include the injection of simulated process signals into the channel as close to the sensors as practicable to verify OPERABILITY including alarm and/or trip functions.

(7) -

Above 70% of RATED THERMAL POWER, verify that the total RCS flow rate as indicated by each CPC is less than or equal to the actual RCS total flow rate determined by either using the reactor coolant pump differential pres-g sure instrumentation (conservatively compensated for measurement uncertain-ties) or by calorimetric calculations (conservatively compensated for measurement uncertainties) and if necessary, adjust the CPC addressable constant flow coefficients such that each CPC indicated flow is less than or equal to the actual flow rate.

The flow measurement uncertainty may be included in the BERR1 term in the CPC and is equal to or greater than 4%.

(8)

Above 70% of RATED THERMAL POWER, verify that the total RCS flow rate as indicated by each CPC is less than or equal to the actual RCS total flow rate determined by calorimetric calculations (conservatively compensated for measurement uncertainties).

(9) -

Above 55% of RATED THERMAL POWER.

(10) -

Detector plateau curves shall be obtained, evaluated, and compared to manufacturer's data.

(11) -

The monthly CHANNEL FUNCTIONAL TEST shall include verification that the correct values of addressable constants are installed in each OPERABLE CPC per Specification 2.2.2.

(12) -

.it least once per 18 months and following maintenance or adjustment of the reactor trip breakers, the CHANNEL FUNCTIONAL TEST shall include independent verification of the undervoltage and shunt trips.

SAN ONOFRE-UNIT 2 3/4 3-12

o INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Feature Actuation System (ESFAS) instrumentation channels and bypasses shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5.

APPLICABILITY:

As shown in Table 3.3-3.*

ACTION:

With an ESFAS instrumentation channel trip setpoint less conservative a.

than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value.

b.

With an ESFAS instrumentation channel inoperable, take the ACTION shown in Table 3.3-3.

SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the per formance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-2.

4.3.2.2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by bypass operation.

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.

4.3.2.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months.

Each test shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the

" Total No. of Channels" Column of Table 3.3-3.

See Special Test Exception 3.10.5 SAN ONOFRE-UNIT 2 3/4 3-13

TABLE 3.3-3 9

Z ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION O

MINIMUM A

TOTAL NO.

CHANNELS CHANNELS APPLICABLE y

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION U

l.

SAFETY INJECTION (SIAS)'

N i

a.. Manual (Trip Buttons) 2 sets of 2 1 set of 2 2 sest of 2 1, 2, 3, 4 8

b., Containment Pressure -

High 4

2 3

1, 2, 3 9*,

10*

c.

Pressurizer Pressure -

Low 4

2 3

1, 2,.3(a) 9*,

10*

d.

Automatic Actuation -

Logic 4

2 3

1, 2, 3, 4 9*,

10*

,s 2.

CONTAINMENT SPRAY (CSAS) w l

)

a.

Manual (Trip Buttons) 2 sets of 2 1 set of 2 2 sets of 2 1, 2, 3 8

l N

y b.

Containment Pressure --

Z High - High 4

2(b) 3 1,2,3 9*, 10*

~

c.

Automatic Actuation Logic 4

2 3

1, 2, 3 9*,

10*

l 3.

CONTAINMENT ISOLATION (CIAS) a.

Manual CIAS (Trip Buttons) 2 sets of 2 1 set of 2 2 sets of 2 1, 2, 3, 4 8

3 y

b.

Manual SIAS (Trip g

Buttons) (c) 2 sets of 2 1 set of 2 2 sets of 2

~l, 2, 3, 4 8

i c.

Containment Pressure -

k High 4

2 3

1, 2, 3 9*,

10*

z d.

Automatic Actuation

?

Logic 4

2 3

1, 2, 3, 4 9*,

10*

a i

j se.

(

TABLE 3.3-3 (Continued) m ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION E

i MINIMUM A

TOTAL NO.

CHANNELS CllANNELS APPLICABLE

.g FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION Q

4.

MAIN STEAM LINE ISOLATION N

a.

Manual (Trip 2/ steam 1/ steam 2/ operating 1,2,3 11 Buttons) generator generator steam i

generator b.

Steam Generator 4/ steam 2/ steam 3/ steam 1, 2, 3 9*,

10*

j Pressure - Low generator generator generator c.

Automatic Actuation 4/ steam 2/ steam 3/ steam 1, 2, 3 9*,

10*

Logic generator generator generator 5.

RECIRCULATION (RAS) w

.l 1

a.

Refueling Water Storage w

Tank - Low 4

2 3

1,2,3,4 9*,

10*

+

h b.

Automatic Actuation Logic 4

2 3

1,2,3,4 9*,

10*

6.

CONTAINMENT COOLING (CCAS) 1 a.

Manual CCAS (Trip Buttons) 2 sets of 2 1 set of 2 2 sets of 2 1,2,3,4 8

1 b.

Manual SIAS (Trip 1

[

Buttons) 2 sets of 2 1 set of 2 2 sets of 2 1,2,3 8

1 5

c.

Automatic Actuation l

Logic.

4 2

3 1,2,3,4 9 *, 10*

r 1

1 eeo e

TABLE 3.3-3 (Continued)

~

1,

[

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTUMENTATION E

j gi MINIMUM c:

TOTAL NO.

CHANNELS CHANNELS APPLICABLE j

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 7.

LOSS OF POWER (LOV) 4 a.

4.16 kv Emergency Bus undervoltage (Loss of' Voltage and

.]

Degraded Voltage) 4/ Bus 2/ Bus 3/ Bus 1, 2, 3, '4 9*, 10*

1-i j

8.

EMERGENCY FEE 0 WATER (EFAS) a.

Manual (Trip Buttons) 2 sets of 2 1 set of 2 2 sets of 2 1,2,3 11 i

per S/G per S/G per S/G u2 l

u

(

u2 b.

Automatic Actuation i. ]:.

O Logic 4/SG 2/SG 3/SG 1, 2, 3 9*, 10*

m c.

SG Level (A/B) - Low and AP (A/B) - High 4/SG 2/SG 3/SG 1, 2, 3 9*, 10

3 d.

SG Level (A/B) - Low l !

and No S/G Pressure -

Low Trip (A/B) 4/SG-2/SG 3/SG 1, 2, 3 9*,

10*

i a

i 4

e hg4

~

TABLE 3.3-3 (Continued)

TABLE NOTATION (a) Trip function may be bypassed in this MODE when pressurizer pressure is less than 400 psia; bypass shall be automatically removed when pressurizer pressure is greater than or equal to 400 psia.

l (b) An SIAS signal is first necessary to enable CSAS logic.

(c) Actuated equipment only; does not result in CIAS.

The provisions of Specification 3.0.3 are not applicable.

The provisions of Specification 3.0.4 are not applicable.

With irradiated fuel in the storage pool.

ACTION STATEMENTS ACTION 8 -

With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 9 -

With the number of channels OPERABLE one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may continue provided the inoperable channel is placed in the bypassed or tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

If the inoperable channel is bypassed, the desirability of maintaining this channel in the bypassed condition shall be reviewed in accordance with Specification 6.5.1.6e.

The channel shall be returned to OPERABLE status no later than during the next COLD SHUTDOWN.

j With a channel process measurement circuit that affects multiple functional units inoperable or in test, bypass or trip all associated functional units as listed.310w.

Process Measurement Circuit Functional Unit Bypassed 1.

Containment Pressure - High Containment Pressure - High (ESF)

Containment Pressure - High (RPS) 2.

Steam Generator Pressure -

Steam Generator Pressure - Low Low Steam Generator AP 1 and 2 (EFAS) 3.

Steam Generator Level Steam Generator Level - Low Steam Generator Level - High Steam Generator AP (EFAS)

SAN ONOFRE-UNIT 2 3/4 3-19 Amendment No. 4

.y...,

.. - -.~

... ~..

~

TABLE 3.3-3.(Continued)

TABLE NOTATION ACTION 10 -

With the number of channels OPERABLE one less than the Minimum Channels OPERABLE, STARTUP and/or POWER OPERATION may continue provided the following conditions are satisfied:

a.

Verify that one of the inoperable channels has been bypassed and place the other inoperable channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , b.

All functional units affected by the bypassed / tripped channel shall also be placed in the bypassed / tripped condition as listed.below:

Process Measurement Circuit Functional Unit Bypassed / Tripped 1.

Containment Pressure Circuit Containment Pressure - High (ESF)

Containment Pressure - High (RPS) 2.

Steam Generator Pressure -

Steam Generator Pressure - Low Low Steam Generator AP 1 and 2 (EFAS) 3.

Steam Generator Level - Low Steam Generator Level - Low Steam Generator Level - High Steam Generator AP (EFAS)

STARTUP and/or POWER OPERATION may continue until the performance of the next required CHANNEL FUNCTIONAL TEST.

Subsequent STARTUP and/or POWER OPERATION may continue if one channel is restored to OPERABLE status and the provisions of ACTION 9 are satisfied.

.t ACTION 11 -

With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channels to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 12 -

With the number of OPERABLE channels one less than the Total Number of Channels, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for surveillance testing provided the other channel is OPERABLE.

{

l l

l SAN ONOFRE-UNIT 2 3/4'3-20 t

l

Table 3.3-3 (Continued)

TABLE NOTATION ACTION 13 With the number of channels OPERABLE less tnan required by the minimum channels OPERABLE requirement, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months initiate and maintain operation of the control room emergency air cleanup system in the emergency (except as required by ACTIONS 14, 15) mode of operation.

ACTION 14 With the number of channels OPERABLE one less than the tntal number of channels, restore the inoperable channel to OPERABLE status within 7 days or-within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months initiate and maintain operation of the control room emergency air cleanup system in the isolation mode of operation.

ACTION 15 With the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months initiate and maintain operation of the control room emergency air cleanup system in the isolation mode of operation.

ACTION 16 With the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.12.

ACTION 17 With the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.9.

(Mode 6 only)

ACTION 17a -

With the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.5.1.

(Mode 1, 2, 3, 4 only)

ACTION 17b -

With the number of channels OPERABLE less than required by the g

minimum channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.3.3.9.

(At all times)

SAN ONOFRE-UNIT 2 3/4 3-21

1 1

i l

TABLE 3.3-4 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES i

S ALLOWABLE E

FUNCTIONAL UNIT TRIP VALUE VALUES a

1.

SAFETY INJECTION (SIAS)

-a j

m a.

Manual (Trip Buttons)

Not Applicable Not Applicable b.

Containment Pressure - High 5 2.95 psig 5 3.14 psig 3

c.

Pressurizer Pressure - Low 2 1806 psia (1) 2 1763 psia (1) d.

Automatic Actuation Logic Not Applicable Not Applicable 2.

CONTAINHENT SPRAY (CSAS) a.

Manual (Trip Buttons)

Not Applicable Not Applicable b.

Containment Pressure -- High-High 5 16.14 psig 5 16.83 psig

)

c.

Automatic Actuation Logic Not Applicable Not Applicable 3.

' CONTAINMENT ISOLATION (CIAS) a.

Manual CIAS'(Trip Buttons)

Not Applicable Not Applicable I

b.

Manual SIAS (Trip Buttons)(5)

Not Applicable Not Applicable c.

Containment Pressure - liigh 5 2.95 psig i 3.14 psig d.

Automatic Actuation Logic Not Applicable Not Applicable 4.

MAIN STEAM ISOLATION (MSIS)

F g

a.

Manual (Trip Buttons)

Not Applicable Not Applicable k

b.

Steam Generator Pressure - Low 2 729 psia (2) 2 711 psia (2) 5 c.

Automatic Actuation Logic Not Applicable Not Applicable 2

5.

RECIRCULATION (RAS)

^

a.

Refueling Water Storage Tank 18.5% of tap span 19.27% 2 tap span 2 17.73%

b.

Automatic Actuation Logic Not Applicable Not Applicable dee

1 INSTRUMENTATION i

ACCIDENT MONITORING INSTRUMENTATION i

LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels.shown in Table 3.3-10 shall be OPERABLE.

APPLICABILITY:

MODES 1, 2 and 3.*

ACTION:

With one or more radiation monitoring alarm channels inoperable, a.

take the ACTION shown in Table 3.3-10.

b.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.6 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 4.3-7.

I*

'l i

1

See Special Test Exception 3.10.5.

L i

i SAN ON0FRE-UNIT 2 3/4 3-51 Amendment No. 4

TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION E

A REQUIRED MINIMUM ch NUMBER OF CHANNELS 5

INSTRUMENT CHANNELS OPERABLE

-ACTION w

N 1.

Containment Pressure - Narrow Range 2

1 20, 21 2.

Containment Pressure - Wide Range 2

1 20, 21 3.

Reactor Coolant Outlet Temperature -

2 1

20, 21 THot (Wide Range) 4.

Reactor Coolant Inlet Temperature -

2 1

20, 21 TCold (Wide Range) l S.

Pressurizer Pressure - Wide Range 2

1 20, 21

'-s 6.

Pressurizer Water Level 2

1 20, 21 Y

7.

Steam Line Pressure 2/ steam 1/ steam 20, 21 generator generator 8.

Steam Generator Water Level - Wide Range 2/ steam 1/ steam 20, 21 generator generator 9.

Refueling Water Storage Tank Water Level 2

1 20, 21 10.

Auxiliary Feedwater Flow Rate 1/ steam N.A.

20 generator 7

11.

Reactor Coolant System Subcooling 2

1 20, 21 Oj Margin Monitor k

12.

Safety Valve Position Indicator 1/ valve N.A.

20 5

13.

Sp' ray System Pressure 2

1 20, 21 E

14.

LPSI Header Temperature 2

1 20, 21

^

15.

Containment Temperature 2

1 20, 21 16.

Containment Water Level - Narrow Range 2

1 20, 21

O Q

TABLE 3.3-10

=

ACCIDENT MONITORING INSTRUMENTATION (CONTINUED)

Si?

?

j_

REQUIRED MINIMUM S

NUMBER OF CHANNELS

-4 INSTRUMENT CHANNELS OPERABLE ACTION ro 17.

Containment Water Level - Wide Range 2

1 20, 21 18.

Core Exit Thermocouples 7/ core 4/ core 20, 21 quadrant quadrant 19.

Containment Area Radiation - High Range 2

1 22, 23 20.

Main Steam Line Area Radiation 1/ steam line N.A.

22 l

21.

Condenser Evacuation System Radiation 1

N.A.

22 u,);

Monitor - Wide Range 12 2.

Purge / Vent Stack Radiation Monitor -

2 1

22, 23

')

Wide Range

23.

Cold Leg HPSI Flow 1/ cold leg N.A.

20 24.

Hot Leg HPSI Flow 1/ hot leg N.A.

20 E

0 NOTES:

The two required channels are the Unit 2 monitor and the Unit 3 monitor.

g.

e...

TABLE 3.3-10 (Continued)

ACTION STATEMENTS ACTION 20 -

With the number of OPERABLE accident monitoring channels less than the Required Number of Channels, either restore the inoperable channel to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 21 -

With the number of OPERABLE accident monitoring channels less than the Minimum Channels OPERABLE requirement, either restore the inoperable channel (s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 22 -

With the number of channels OPERABLE less than the Required N:smber of Channels, comply with the ACTION requirements of Specification 3.3.3.6.

ACTION 23 -

With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirement, either restore the inoperable Channel (s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or:

1)

Initiate-the preplanned alternate method of monitoring the appropriate parameter (s), and 2)

Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

I SAN ON0FRE-UNIT 2 3/4 3-53a Amendment No. 4

+

3/4.~4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1 Both Reactor Coolant loops and both Reactor Coolant pumps in each loop l

shall be in operation.

APPLICABILITY: 1 and 2.*

ACTION:

With less than the above required Reactor Coolant pumps in operation, be in at least HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

A SURVEILLANCE REQUIREMENTS 4.4.1.1 The above required Reactor Coolant loops shall be verified to be in l

operation and circulating Reactor Coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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See Special Test Exception 3.10.3.

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l SAN ONOFRE-UNIT 2 3/4 4-1 Amendment No. 4

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REACTOR COOLANT SYSTEM

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HOT STANDBY LIMITING CONDITION FOR OPERATION

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3.4.1.2 a.

The Reactor Coolant loops listed below shall be OPERABLE:

1.

Reactor Coolant Loop 1 and its associated steam generator and at least 'one associated Reactor Coolant pinp.

2.

Reactor Coolant. Loop 2 and its associated steam generator and at least one associated Reactor Coolant pump.

b.

At least one of the above Reactor Coolant loops shall be in operation.*

~

APPLICABILITY:

MODE 3

\\

s ACTION:

i With less than the;above required Reactor Coola'nt loops a.

OPERABLE, restore the regiired loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTOOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

With no Reacto'r Coolant loop in operation, suspend all operations involving a reduction in baron concentration of the Reactor Coolant System and immediately, initiate corrective action to return the required Reactor Coo ant'.' loop to operation.

SURVEILLANCE REQUIREMENTS f.

4.4.1.2.1 At least the above required Reactor Coolant pumps, if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4.1.2.2 At least one Reactor Coolant loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hoces.

4.4.1.2.3 The required steam generator (s) shall be determined OPERABLE verifying the secondary side water leJel to be > 10% (wide range) at least once y

per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

A s

All Reactor Coolant pumps may be de energized /c. g I hour provided (1) no operations are permitted that would cauts diluu u of the Reactor Coolant System baron concentration, and (2) core outlet temperature is s

maintained at least 10*F below saturation temperature.

SAN ONOFRE-UNIT 2 3/4 4-2

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[EACTOR COOLANT SYSTEM HOT SHUTDOWN x < LIMITING CONDITION FOR OPERATION

,q, 3.4.1.3 a.

At least two of the loop (s)/ train (s) listed below shall be s

OPERABLE:

1.

Reactor Coolant Loop 1 and its associated steam generator and at least one associated Reactor Coolant pump,**

w 2.

Reactor Coolant Loop 2 and its associated steam generator and at least one associated Reactor Coolant pump,**

3.

Shutdown Cooling Train A, 4.

Shutdown Cooling Train B.

b.

At least one of the above Reactor Coolant loops and/or shutdown cooling trains shall be in operation.*

APPLICABILITY:

MODE 4 ACTION:

a.

With less than the above required Reactor Coolant loops and/or shutdown cooling trains OPERABLE, immediately initiate correc-tive action to return the required loops / trains to OPERABLE status as soon as possible; if the remaining OPERABLE loop is a shutdown cooling train, be in COLD SHUTDOWN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

With no Reactor Coolant loop or shutdown cooling train in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immedi-ately initiate corrective action to return the required coolant loop / train to operation.

A All Reactor Coolant pumps and shutdown cooling pumps may be de-energized for up to I hour provided (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10 F below saturation temperature.

- A Reactor Coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 235 F unless

1) the pressurizer water volume is less than 900 cubic feet or 2) the secondary water temperature of each steam generator is less than 100 F above each of the Reactor Coolant System cold leg temperatures.

l SAN ON0FRE-UNIT 2 3/4 4-3 Amendment No. 4

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REACTOR COOLANT SYSTEM HOT SHUTOOWN SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required Reactor Coolant pump (s), if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct. breaker alignments and indicated power availability.

4.4.1.3.2 The required steam generator (s) shall be determined OPERABLE by verifying the secondary side water level to be > 10% (wide range) at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.3.3 At least one Reactor Coolant loop or shutdown cooling train shall be verified to be in operation and circulating Reactor Coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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l SAN ON0FRE-UNIT 2 3/4 4-4

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REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4.1 At least one shutdown cooling train shall be OPERABLE and in operation,* and either; One additional shutdown cooling train shall be.0PERABLE,#

a.

or b.

The secondary side water level of each steam generator shall be greater than 10% (wide range).

APPLICABILITY: MODE 5, with Reactor Coolant loops filled.

ACTION:

With less than the above required shutdown trains / loops OPERABLE or a.

with less than the required steam generator level, immediately initiate corrective action to return the required trains / loops to OPERABLE status or restore the required level as soon as possible.

b.

With no shutdown cooling train in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required shutdown cooling train to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.4.1.1 The secondary side water level of at least two steam generators, when required, shall be determined to be within limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.4.1.2 The shutdown cooling train shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

0ne shutdown cooling train may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other shutdown cooling train is OPERABLE and in operation.

n The shutdown cooling pump may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided

1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and 2) core outlet temperature is maintained at least 10 F below saturation temperature.

SAN ONOFRE-UNIT 2 3/4 4-5 Amendment No. 4 v---

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REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS NOT FILLED i

LIMITING CONDITION FOR OPERATION 3.4.1.4.2 Two shutdown cooling trains shall be OPERABLE # and at least one shutdown cooling train shall be in operation.*

APPLICABILITY:

MODES 5 with Reactor Coolant loops not filled.

ACTION:

With less than the above required trains OPERABLE, immediately a.

initiate corrective action to return the required trains to OPERABLE status as soon as possible.

b.

With no shutdown cooling trains in operation, suspend all operations involving a reduction in baron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required shutdown cooling train to operation.

SURVEILLANCE REQUIREMENTS t

4.4.1.4.2 At least one shutdown cooling train shall be determined to be in oper.ition and circulating reactor coolant at least'once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

0ne shutdown cooling train may be inoperable'for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other shutdown cooling train is OPERABLE and in operation.

x The shutdown cooling pump may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided 1) no operations are nermitted that would cause dilution of the Reactor Coolant System boron conc.ntration, and 2) core outlet temperature is maintained at least 10 F below saturation temperature.

SAN ONOFRE-UNIT 2 3/4 4-6

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REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES - OPERATING LIMITING CONDITION FOR OPERATION 3.4.2 All pressurizer code safety valves shall be OPERABLE with a lift

. setting of 2500 PSIA i 1%.*

APPLICABILITY:

MODES 1, 2 and 3.

ACTION:

With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.2 No additional Surveillance Requirements other than those required by Specification 4.0.5.

AThe lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

SAN ONOFRE-UNIT 2 3/4 4-7

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REACTOR COOLANT SYSTEM 3/4.4.3 PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.3 The pressurizer shall be OPERABLE with a water volume of less than or equal to 900 cubic feet and at least two groups of pressurizer heaters powered from the 1E busses, each having a capacity of at least 150 kw.

APPLICABILITY:

MODES 1, 2 and 3.

ACTION:

a.

With one group of pressurizer heaters inoperable, restore at least two groups to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , b.

With the pressurizer otherwise inoperable, 'be in at least HOT STANDBY with the reactor trip breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 5

4.4.3.1 The pressurizer water volume shall be determined to be within its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.3.2 The pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by manually energizing the heaters.

4.4.3.3 The capacity of each of the above required groups of pressurizer heaters

~ hall be verified by energizing the heaters and measuring circuit current at s

least once per 92 days.

SAN ON0FRE-UNIT 2 3/4 4-8 Amendment No. 4

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REACTOR COOLANT SYSTEM

. PRESSURIZER - HEATUP/COOLDOWN l

LIMITING CONDITION FOR OPERATION 3.4.8.2 The pressurizer shall be limited to:

a.

A maximum heatup of 200 F in any one hour period, b.

A maximum cooldown of 200 F in any one hour period.

APPLICABILITY:

At all times.

1 ACTION:

With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within-30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition' on the structural integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the pressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.8.2.1 The pressurizer temperatures shall be determined'to be within the i

limits at least once per 30 minutes during system heatup or cooldown.

4.4.8.2.2 The spray water temperature differential shall be determined for use in Table 5.7-1 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during auxiliary spray operation.

t SAN ONOFRE-UNIT 2 3/4 4-31 Amendment No. 4

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REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS RCS TEMPERATURE I 235*F LIMITING CONDITION FOR OPERATION 3.4.8.3.1 At least one of the following overpressure protection systems shall be OPERABLE:

a.

The Shutdown Cooling System Relief Valve (PSV9349) with:

1)

A lift setting of 406 1 10 psig*, and 2)

Relief Valve isolation valves 2HV9337, 2HV9339, 2HV9377 and 2HV9378 open, or, b.

The Reactor Coolant System depressurized with an RCS vent of greater than or equal to 5.6 square inches.

APPLICABILITY:

MODE 4 when the temperature of any one RCS cold leg is less l

than or equal to 235 F; MODE 5; MODE 6 with the reactor vessel head on.

ACTION:

a.

With the SDCS Relief Valve inoperable, reduce T to less than avg 200 F, depressurize and vent the RCS through a greater than or equal to 5.6 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b.

With one or both SDCS Relief Valve isolation valves in a single SDCS Relief Valve isolation valve pair (valve pair 2HV9337 and 2HV9339 or valve pair 2HV9377 and 2HV9378) closed, open the closed valve (s) within 7 days or reduce T to less than 200 F, depres-avg surize and vent the RCS through a greater than or equal to 5.6 inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

c.

In the event either the SDCS Relief Valve or an RCS vent is used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days.

The report shall describe the circumstances initi-ating the transient, the effect of the SDCS Relief Valve or RCS vent I

on the transient and any corrective action necessary to prevent l

recurrence.

I d.

The provisions of Specification 3.0.4 are not applicable.

l SURVEILLANCE REQUIREMENTS i

4.4.8.3.1.1 The SDCS Relief Valve shall be demonstrated OPERABLE by:

a.

Verifying at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when the SDCS Relief Valve is being used for overpressure-protection that SDCS Relief Valve i

l isolation valves 2HV9337, 2HV9339, 2HV9377 and 2HV9378 are open.

t For valve temperatures less than or equal to 130 F.

l l

SAN ON0FRE-UNIT 2 3/4 4-32 Amendment No. 4 l

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REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS i

RCS TEMPERATURE > 235*F LIMITING CONDITION FOR OPERATION

~3.4.8.3.2 At least one of the following overpressure protection systems shall be OPERABLE:

a.

The Shutdown Cooling System Relief Valve (PSV9349) with:

1)

A lift setting of 406 i 10 psig*, and 2)

Relief Valve isolation valves 2HV9337, 2HV9339, 2HV9377 and 2HV9378 open, or, b.

A minimum of one pressurizer code safety valv.e with a lift setting of 2500 psia + 1%**.

APPLICABILITY: MODE 4 with RCS temperature above 235*F.

ACTION:

With no safety or relief valve OPERABLE, be in COLD SHUTDOWN and vent the a.

RCS through a greater than or equal to 5.6 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b.

In the event the SDCS Relief Valve or an RCS vent is used to mitigate an RCS pressure transient-a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days.

The report shall describe the circumstances initiating the transient, the effect of the SDCS Relief Valve code safety valve or RCS vent on the transient and any corrective action necessary to prevent recurrence.

SURVEILLANCE REQUIREMENTS j

4.4.8.3.2.1 The SDCS Relief Valve shall be demonstrated OPERABLE by:

a.

Verifying at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months that the SDCS Relief Valve isolation valves 2HV9337, 2HV9339, 2HV9377 and 2HV9378 are open when the SDCS Relief Valve is being used for overpressure protection.

b.

Verifying relief valve setpoint at least once per 30 months when tested pursuant to Specification 4.0.5.

4.4.8.3.2.2 The pressurizer code safety valve has no additional surveillance requirements other than those required by Specification 4.0.5.

4.4.8.3.2.3 The RCS vent shall be verified to be open at.least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the vent is being used for overpressure protection, except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.

"For valve temperatures less than or equal to 130 F.

- The litt setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

SAN ONOFRE-UNIT 2 3/4 4-33 Amendment No. 4

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O REACTOR COOLANT SYSTEM 3.4.9 STRUCTURAL INTEGRITY l

LIMITING CONDITION FOR OPERATION 3.4.9 The structural integrity of ASME Code Class 1, 2 and 3 components shall be maintained in accordance with Specification 4.4.9.

APPLICABILITY: ALL MODES ACTION:

a.

With the structural integrity of any ASME Code Class 1 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50 F above the minimum temperature required by NDT considerations.

b.

With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200 F.

c.

With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural integrity of the affected component to within its limit or isolate the affected component from service.

d.

The provisions of Specification 3.0.4 are not applicable.

3 SURVEILLANCE REQUIREMENTS 4.4.9 In addition to the requirements of Specification 4.0.5, each Reactor Coolant Pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975.

i SAN ONOFRE-UNIT 2 3/4 4-34

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CONTAINMENT SYSTEMS i

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CONTAINMENT AIR LOCKS l

LIMITING CONDITION FOR OPERATION

3. 6.1. 3 Each containment air lock shall be OPERABLE with:

a.

Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and b.

An overall air lock leakage rate of less than or equal to 0.05 L at a

P,, (55.7 psig).

APPLICABILITY:

MODES 1, 2, 3 and 4.

ACTION:

a.

With one containment' air lock door inoperable:

1.

Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the OPERABLE air lock door closed.

i 2.

Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days.

3.

Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

4.

The provisions of Specification 3.0.4 are not applicable.

b.

With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

l SAN _0NOFRE-UNIT 2 3/4 6-5 i

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

Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following each closing, except'when the air lock is a.

~

being used for multiple entries, then at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , by verifying that seal leakage is less than or equal to 0.01 La when determined by flow measurement, with the volume between the door seals pressurized to greater than or equal to 9.5 0.5 psig for at least 15 minutes, b.

By conducting overall air lock leakage tests at not less than P (55.7 psig), and verifying the overall air lock leakage rate is within its limit:

1.

At least once per 6 months, and 2.

Prior to establishing CONTAINMENT INTEGRITY when maintenance has been performed on the air lock that could af fect the air lock sealing capability.*

i c.

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

1 a

The provisions of Specification 4.0.2 are not applicable.'

Exemption to Appendix J of 10 CFR 50.

s i

i SAN ONOFRE-UNIT 2 3/4 6-6 Amendment No. 4 4

CONTAINMENT SYSTEMS CONTAINMENT STRUCTURAL INTEGRITY 3

LIMITING CONDITION FOR OPERATION 3.6..l.6 The structural integrity of the containment shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6.

APPLICABILITY:

MODES 1, 2, 3 and 4.

ACTION:

With the structural integrity of the containment not conforming to the above requirements, perform an engineering evaluation of the containment to demonstrate its structural integrity within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; otherwise, be in at least HOT STANDBY

~

within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

~

SURVEILLANCE REQUIREMENTS 4.6.1.6 Containment Tendons The containment's structural integrity shall be demonstrated at the end of one, three and five years after the initial structural integrity test (ISIT) and every five years thereafter with the exception of tendon lift off force and tendon detensioning and material tests and inspections which shall be determined at the end of one, five and ten years following the ISIT and every ten years thereafter in accordance with Table 4.6-1.

The structural integrity shall be demonstrated by:

a.

Determining that tendons selected in accordance with Table 4.6-1 have a lift off force between the maximum and minimum v'alues listed 2

in Table 4.6-2 at the first year inspection.

For subsequent inspec-tions, for tendons and periodicities per Table 4.6-1, the maximum first year lif t off forces shall be decreased by the amount X1 log t kips for U tendons and Y1 log t kips for hoop tendons and the minimum lift off forces shall be decreased by the amount X2 log t for U tendons and Y2 log t for hoop tendons where t is the time interval in years from initial tensioning of the tendon to the current testing date and the values X1, X2, Y1 and Y2 are in accordance with the values listed in Table 4.6-2 for the surveillance tendon.

This test shall include essentially a complete detensioning of tendons selected in accordance with Table 4.6-1 in which the tendon is detensioned to determine if any wires or strands are broken or damaged.

Tendons found acceptable during this test shall be retensioned to obtain a lift off force equal to +0, -5% of the prescribed upper limit.

During retensioning of these tendons, the change in load and elongation shall be measured simultaneously at a minimum of three, approximately equally spaced, levels of force between the seating force and zero.

If elonga-tion corresponding to a specific load differs by more than 5% from that recorded during installation of tendons, an investigation should SAN ON0FRE-UNIT 2 3/4 6-9 Amendment No. 4

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CONTAINMENT SYSTEMS 1

SURVEILLANCE REQUIREMENTS (Continued)

]

be made to ensure that such difference is not related to wire failures or slip of wires in anchorages.

If the lift off force of-any one tendon in the total sample population lies between the i

prescribed lower limit and 90% of the prescribed lower limit, two tendons, one on each side of this tendon shall be checked for their lift off force.

If both of these adjacent tendons are found acceptable, the surveillance program may proceed considering the single deficiency as unique and acceptable.

The tendon (s) shall be restored to the required level of integrity.

More than one tendon below the predicted bounds out of the original sample population or the lift off force of a selected tendon lying below 90% of the prescribed lower limit is evidence of abnormal degradation of the containment structure.

b.

Performing tendon detensioning and material tests and inspections of a previously stressed tendon wire or strand from one tendon of each group (hoop and U), and determining over the entire length of the removed wire or strand that:

1.

The tendon wires or strands are free of corrosion, cracks and damage.

1 1

2.

A minimum tensile strength value of 270 ksi (guaranteed ultimate l

strength of the tendon material) for at least three wire or strand samples (one from each end and one at mid-length) cut from each removed wire or strand.

Failure of any one of the wire or strand samples to meet the minimum tensile strength test is evidence of abnormal degradation of the containment i

structure.

Performing a visual inspection of the following:

c.

)

1.

Containment Surfaces - The structural integrity of the exposed i

accessible interior and exterior surfaces of the containment shall be determined during the shutdown for, and prior to, each Type A containment leakage rate test (Specification 4.6.1.2) by a visual inspection of these surfaces and verifying no apparent changes in appearance or other abnormal degradation (e.g.,

I widespread cracking, spalling and/or grease leakage).

2.

End Anchorages - The structural integrity of the end anchorages ~

(e.g., bearing plates, stressing washers, shims, wedges and anchorheads) of all tendons inspected pursuant to Specification 4.6.1.6a shall be demonstrated by inspection that no apparent changes have occurred in the visual appearance of the end anchorage.

3.

Concrete Surfaces - The structural integrity of the concrete surfaces adjacent to the end anchorages of tendons inspected I

SAN ONOFRE-UNIT 2 3/4 6-10 Amendment No. 4

CGNTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) pursuant to Specification 4.6.1.6a shall be demonstrated by visual examination of the crack patterns to verify no abnormal material behavior.

d.

Verifying the OPERABILITY of the sheathing filler grease by the following:

1.

No significant voids (in excess at 5% of the net duct volume),

or the presence of free water, within the grease filler material, taking into account temperature variations.

2.

No significant changes have occurred in the physical appearance of the sheathing filler grease.

3.

Complete grease coverage exists for the anchorage system.

4.

Chemical preperties are within the tolerance limits specified by the sheoching filler grease manufacturer.

l

.i.

i I

SAN ONOFRE-UNIT 2 3/4 6-11 Amendment No. 4 4

~T

'*?'~

.,,,,.,,, ~ ~

R ^., ",

TABLE 4.6-1 TENDON SURVEILLANCE TENDON NUMBERS Years Af ter Initial Structural 1

3 5

10 15 Integrity Test Type of Inspection H

U H

U Visual Inspection 20 31-121 5

13-139 42 64-178 of End Anchorages 86 9-143 36 35-117 86 9-143 20-66-176 50 12-140 and Adjacent 97 66-176 79 4-58 75 94-148 86 9-143 114 5-57 Concrete Surface 53 88-154 113 78-164 9

19-133 53 39-113 13 96-146 64 A7 108 20 31-121 42 64-178 Prestress 86 9-143 86 9-143 20 66-176 Monitoring 97 66-176 75 94-148 86 9-143 Tests 53 88-154 9

19-133 53 39-113 64

'na Dentensioning and 97 88-154 42 19-133 20 66-176 Material Tests TENDON NUMBERS Years Af ter Initial Structural 20 25 30.

35 40 I

Intecritv Test Type of Inspection H

U H

U Visual Inspection of End Anchorages 75 86-156 12 24-128 86f 9-143 81 41-111 97 9-143 and Adjacent 86 9-143 90 70-172 31 69-178 109 90-152 86 31-121 i

Concrete Surface 9

43-109 25 76-166 64 94-148 31 50-102 LO8 86-156 Prestress 7S'86-156 86 9-143 97 9-143 Monitoring 86 9-143 31 64-178' 86 31-121 Tests 9

43-109 64 94-148 108 86-156 Detensioning and 75 43-109 31 64-178 86 9-143 Material Tests SAN ONOFRE-UNIT 2 3/4 6-12

l a

i TABLE 4.6-2 TENDON LIFT-OFF FORCE 2

U TENDONS i

Tendon First Year Number Ends Maximum (kips)

Minimum (kips)

X1 X2 43-109 43 1634 1457 21.2 31.2 109 1604 1431 20.6 30.0 39-113 39 1625 1449 21.8 31.8 113 1601 1428 20.0 30.0 31-121 31 1574 1406 21.2 29.3 121 1586 1415 21.2 30.0 19-133 19 1644 1465 22.5 31.3 133 1593 1423 20.6 30.0

)

9-143 9

1618 1444 21.8 31.2 143 1598 1428 20.6 30.0 i

94-148 94 1560 1394 19.4 29.3 14P; 1570 1403 20.6 28.-7 88-154 88 1588 1415 21.2 30.0 154 1568 1399 19.4 28.7 86-156 86 1567 1400 20.6 30.0-156 15.~f.

1399 19.4 28.7 66-176 66 lb77 1407 20.6 30.0 176 1579 1409 20.0 30.0 64-178 64 1560 1393 20.0 28.1 178 1582 1412 20.6 28.7 HOOP TENDONS 1

Tendon First Year i

Number Buttress Maximum (kips)

Minimum (kips)

Y1 Y2 i

9 2

1528 1348 26.8 36.8 3

1502 1328 25.6 31.8 20 1

1569 1383 28.1 39.3 3

1527 1348 25.6 36.2 31 1

1443 1281 23.1 31.8 2

1502 1349 24.3 46.2 j

42 2

1577 1398 26.2 36.2 3

1549 1395 24.3

,46.2 53 1

1597 1416 26.2 36.2 3

1564 1390 25.6 35.0 64 1

1607 1426 26.2 37.5 2

1570 1396 25.6 35.6.

75 2

1553 1374 26.2 36.2 3

1525 1371 24.3 35.6 86 1

1600 1423 21.2 31.2 3

1527 1362 20.6 29.3 97 1

1563 1393 20.6 29.3 2

1546 1380 19.4 29.3 108 2

1626 1450 21.8 30.6 3

1587 1418

'20.6 28.7 SAN ONOFRE-UNIT 2 3/4 6-12a Amendment No. 4

O

~

s TABLE 3.6-1 (Continued) z Oz Q

MAXIMUM gj PENETRATION ISOLATION NUMBER VALVE NUMBER FUNCTION TIME (SEC)

-4 D.

OTHER m

3 3"-018-A-551#

High pressure safety injection NA 3

HV-9323#

High pressure safety injection NA 3

HV-9324#

High pressure safety injection NA 5

3"-019-A-551#

High pressure safety injection NA 5

HV-9326#

High pressure safety injection NA 5

HV-9327#

High pressure safety injection NA 8

2"-122-C-554 Charging line to re9enerative heat exchanger NA

{

9 PSV-9349#

Shutdown cooling relief NA l

.11 3"-236-C-675 Demineralized water to service stations and sump pump NA R2 14 4"-061-C-681 Fire protection NA 17 HV-4058#*

Steam generator secondary coolant sample NA T

20 2"-573-C-611 Quench tank makeup NA y

21 2"-017-C-627 Service air supply line NA 22 1-1/2"-016-C-617 Instrument air supply line NA 23A 3/4"-002-C-611

.LP N to containment NA 2

32 HV-8421#

Mainsteam atmospheric dump NA 32 PSV-8410#

Mainsteam relier NA 32 PSV-8411#

Mainsteam relief NA 32 PSV-8412#

Mainsteam relief NA 32 PSV-8413#

Mainsteam relief NA jf 32 PSV-8414#

Mainsteam relief NA g

32 PSV-8415#

Mainsteam relief NA

((

32 PSV-8416#

Mainsteam relief NA g

32 PSV-8417#

Mainsteam relief NA r+

32 PSV-8418#

Mainsteam relief NA gf 32 HV-82498#

Mainsteam trap isolation NA 32 HV-8202#

Mainsteam isolation bypass NA

=

32 HV-8200#

Nainsteam to auxiliary feedwater turbine NA 33 HV-8419#

Mainsteam atmospheric dump NA 33 PSV-8401#

Mainsteam relief NA

. (

TABLE 3.6-1 (Continued)

E O

MAXIMUM h

PENETRATION ISOLATION g

NUMBER VALVE NUMBER FUNCTION TIME (SEC)

Z 33 PSV-8402#

Mainsteam relief NA 33 PSV-8403#

Mainsteam relief NA 33 PSV-8404#

Mainsteam relief NA l

33 PSV-8405#

Mainsteam relief NA 33 PSV-8406#

Mainsteam relief NA 33 PSV-8407#

Mainsteam relief NA 33 PSV-8408#

Mainsteam relief NA 33 PSV-8409#

Mainsteam relief NA l

33 IIV-82488#

Mainsteam trap isolation NA 33 HV-8203#

Mainsteam isolation bypass NA 33 HV-8201#

Mainsteam to auxiliary feedwater turbine NA gg 36 HV-4054#*

Steam generator blowdown NA 37 HV-4053#*

Steam generator blowdown NA

[

39 3"-020-A-551#

High pressure safety injection NA 39 IIV-9329#

High pressure safety injection NA 39 HV-9330#

High pressure safety injection NA 41 3"-021-A-551#

liigh pressure safety injection NA 41 IIV-9332#

liigh pressure safety injection NA 41

.HV-9333#

liigh pressure safety injection NA 42 HV-6223 Component cooling water inlet NA 43 HV-6236 Component cooling water inlet NA 44 HV-4057#*

Steam generator secondary coolant sample NA 48 8"-072-A-552#@

Low pressure safety injection NA 48 HV-9322#9 Low pressure safety injection NA 49 8"-073-A-552#@

Low pressure safety injection NA 49 HV-9325#@

Low pressure safety injection NA 50 8"-074-A-552#@

Low pressure safety injection NA 50 HV-9328#@

Low pressure safety injection NA 51 8"-075-A-552#@

Low pressure safety injection NA 51 liv-9331#@

Low pressure safety injection NA 52 8"-004-C-406 Containment spray inlet NA 52 HV-9367 Containment spray inlet NA

TABLE 3.7-2 MAXIMUM ALLOWABLE LINEAR POWER LEVEL-HIGH TRIP SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING OPERATION WITH BOTH STEAM GENERATORS A-i a

Maximum Allowable Linear Power Z

Maximum Number of Inoperable Safety Level-High Trip Setpoint

-[

Valves on Any Operating Steam Generator (Percent of RATED THERMAL POWER) 1 98.9 2

86.6 3

74.2 R

4 61.8 Y

w I

I

PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:

Two motor-driven auxiliary feedwater pumps, each capable of being a.

powered from separate emergency busses, and b.

One steam turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system.

4 APPLICABILITY:

MODES 1, 2 and 3.

ACTION:

With one auxiliary feedwater pump inoperable, restore the required a.

4 auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With two auxiliary pumps inoperable, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible.

SURVEILLANCE REQUIREMENTS 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE:

i a.

At least once per 31 days by:

1.

Testing the turbine driven pump and both motor driven pumps pursuant to Specification 4.0.5.

The provisions of.

Specification 4.0.4 are not applicable for the turbine driven pump for entry into MODE 3.

i 2.

Verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, ' sealed, or otherwise secured in position, is in its correct position.

3.

Verifying that both manual valves in the suction lines from the primary AFW supply tank (condensate storage tank T-121) to each AFW pump', and the manual discharge line valve of each AFW pump are locked in the open position.

SAN ON0FRE-UNIT 2 3/4 7-4 Amendment No. 4

t i

i i-4 PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 1

3.7.1.5 Each main steam line isolation valve shall be OPERABLE.

j APPLICABILITY: MODES 1, 2 and 3.

ACTION:

i

. MODE 1 With one main steam line isolation valve inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; otherwise, reduce power to less than or equal to 5 percent RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

i MODES 2 With one main steam line isolation valve inoperable, and 3 subseqent operation in MODES 2 or 3 may proceed provided:

a.

The isolation valve is maintained closed.

b.

The orovisions of Specification 3.0.4 are not applicable.

Otherwise, be in at least HOT STANDBY within the'next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months l

and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS

4. 7.1. 5 Each main steam line isolation valve shall be demonstrated'0PERABLE by verifying full closure within 5.0 seconds when tested pursuant to Specification 4.0.5.

I i

i 4

e i

i SAN ONOFRE-UNIT 2 3/4 7-9 Amendment No. 4

~--

PLANT SYSTEMS 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION i

LIMITING CONDITION FOR OPERATION 3.7.2 The temperatures of both the primary and _ secondary coolants in the steam generators shall be greater than 70'F when the pressure of either cool-ant in the steam generator is greater than 200 psig.

APPLICABILITY:

At all times.

ACTION:

With the requirements of the above specification not satisfied:

Reduce the steam generator pressure of the applicable' side to less a.

than or equal to 200 psig within 30 minutes, and b.

Perform an engineering evaluation to determine the effect of the overpressurization on the structural integrity of the steam generator.

Determine that the steam generator remains acceptable for continued operation prior to increasing its temperatures above 200 F.

SURVEILLANCE REQUIREMENTS t

4.7.2 The pressure in each side of the steam generators shall be determined to be less than 200 psig at least once per hour when the temperature of either the primary or secondary coolant is less than 70 F.

l i

SAN ONOFRE-UNIT 2 3/4 7-10 i

O 8

ELECTRICAL POWER SYSTEMS MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION BYPASS LIMI_ TING CONDITION FOR OPERATION 3.8.4.2 The thermal overload protection shall be bypassed by a bypass device integral with the motor starter of each alve listed in Table 3.8.2.

APPLICABILITY:

Whenever the motor operated valve is required to be OPERABLE.

ACTION:

With the thermal overload grotection not bypassed by the integral bypass device, bypass the thermal overload protection within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or declare the affected valve (s) inoperable and apply the appropriate ACTION Statement (s) for the affected valve (s).

SURVE;!'ANCE REQUIREMENTS 4.8.4.2 The above required thermal overicad protection shall be verified to be bypassed by integral bypass devices:

a.

At least once per 18 months, b.

Following maintenance on the valve motor starter, and t

Following any periodic testing during which the thermal overload c.

device was temporarily placed in force.

SAN ONOFRE-UNIT 2 3/4 8-31

'i

--e-

4 TABLE 3.8-2 MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION BYPASS DEVICES Permanently Bypassed VALVE NUMBER FUNCTION HV-9339 Shutdown cooling flow from reactor coolan; loop 2 HV-9340 SI tank T008 to reactor coolant loop 1A.

HV-9370 SI tank T010 to reactor coolant loop 2B HV-9347 SI pump minimum recirculation HV-9322 LPSI Header to reactor coolant loop 1A HV-9331 LPSI Header to reactor coolant loop 2B HV-9348 SI pump minimum recirculation HV-9323 HPSI Header #2 to reactor coolant loop 1A HV-9332 HPSI Header #2 to reactor coolant loop 2B 1

HV-9217 RCP bleed off to volume control tank cont. isol.

HV-9326 HPSI Header #2 to reactor coolant loop 1B HV-9329 HPSI Header #2 to reactor coolant loop 2A HV-7258 Waste gas surge tank header containment isolation HV-0508 Reactor coolant hot leg #1 sample containment isolation HV-0517 Reactor coolant hot leg #2 sample containment isolation HV-9368 Shutdown HX E003 to containment spray Header #2 HV-0510 Pressurizer vapor sample containment isolation HV-0512 Pressurizer surge line liquid sample containment isolation HV-9950 Containment purge outlet to exhaust unit A060 - cont. isol.

HV-9917 Hydrogen purge exhaust' unit A082 inlet - containment isol.

HV-9946 Hydrogen purge supply unit A080 discharge - containment isol.

SAN ONOFRE-UNIT 2 3/4 8-32 Amendment No. 4 w

e

TABLE 3.8-2 (Continued)

VALVE j

NUMBER FUNCTION HV-9302 Containment emergency sump outlet HV-9304 Containment emergency sump outlet HV-6211 CCW Non-Critical Loop to containment - isolation valve HV-6368 CCW Loop B to emergency cooling unit E400 HV-6369 CCW from emergency cooling unit E400 to loop B HV-6216 CCW Non-Critical Loop from containment - isolation valve HV-6372 CCW Loop B to emergency cooling unit E402 HV-6373 CCW Loop B from emergency cooling unit E402 HV-9900 Containment normal cooling supply isolation HV-9971 Containment normal cooling return isolation LV-0227C Boric Acid makeup control HV-4713 Aux. F.W. P141 discharge to steam generator E089 control valve HV-9334 SI tank drain Header to refueling water tank - cont, isol.

HV-9350 SI tank T007 to reactor. coolant loop 1B HV-9360 SI tank T009 to reactor coolant loop 2A HV-9325 LPSI Header to reactor coolant loop 1B HV-9328 LPSI Header to reactor coolant loop 2A HV-9201 Aux. spray to pressurizer HV-9327 HPSI Header #1 to reactor coolant. loop 1B HV-9330 HPSI Header #1 to reactor coolant loop 2A HV-6223 CCW Non-Crit Loop Containment inlet isolation HV-9324 HPSI Header #1 to reactor coolant loop 1A HV-9333 HPSI Header #1 to reactor coolant loop 28 HV-9337 Shutdown coolant flow from reactor coolant loop 2 HV-9377 Shutdown coolant ficw from reactor coolant loop 2 SAN ON0FRE-UNIT 2 3/4 8-33 Amendment No. 4

r:

TABLE 3.8-2 (Continued)

VALVE NUMBER FUNCTION HV-9378 Shutdown coolant flow from reactor coolant loop 2 HV-0516 Reactor coolant drain tank sample containment isolation HV-7512 Containment isolation reactor coolant drain tank to R.W. system HV-9367 Shutdown HX E004 to containment spray header #1 HV-0514 Quench tank vapor sample containment isolation HV-5803 Containment sump to R.W. sump HV-9949 Containment purge inlet from supply unit A374 isol.

HV-9303 Containment emergency sump outlet HV-9305 Containment emergency sump outlet HV-6366 CCW Loop A fo emergency cooling unit E401 HV-6367 CCW Loop A from emergency cooling unit E401 HV-6236 CCW Non-crit. containment outlet isolation valve HV-6370 CCW Loop A to emergency cooling unit E399 HV-6371 CCW Loop A from emergency cooling unit E399 HV-8150 Shutdown cooling HX E004 outlet isolation valve b

HV-8151 Shutdown cooling HX E003 outlet. isolation valve HV-9306 SI pump minimum recirculation HV-9307 SI pump minimum recirculation HV-9247 Boric acid pumps to CVC charging pump suction HV-9379 Shutdown cooling flow to LPSI HV-9353 Shutdown cooling warm-up valve

(

HV-9420 HPSI Header #1 to reactor coolant loop 2 hot leg HV-6497 Saltwater from CCW HX-E001 HV-9300 Refueling water tank east (T-005) outlet i

l SAN ONOFRE-UNIT 2 3/4 8-34 Amendment No. 4

.. - ~.... _..

a

=

TABLE 3.8-2 (Continued)

[

VALVE NUMBER FUNCTION U

HV-5686 Firewater to containment isolation

[..

s HV-02278 Volume control tank (i077) drain return i

HV-9240 Boric acid makeup tank (T071) to charging pump suctio C s

HV-9235 Boric' acid makeup tank (T072) to charginp. pump suctien i

HV-9336 Shutdown cooling flow to LPSI pump suction s

',\\

s HV-9359 Shutdown cooling warm up valve

)

HV-9301 Refueling wa'ter tank west (T-006) outlet,

1 HV-6495 Saltwater from CCW HX-E002 2

TV-9267 Letdown line containment isolation valve HV-9434 HPSI Header #2 to reactor coolant loop 1 hot leg

)

HV-8152 Shutdown cooling HX inlet isolation valve

-2 HV-8153 Shutdown cooling HX inlet isolation valve '

i N

HV-4712 Aux F.W. pump P504 discharge to Steam gen. control Q

.t I

,s

. s s

N

\\

d i

s' i

,,. 4 SAN ONOFRE-UNIT'2 3/4 E.-35

/

Amen 3 den t l'!o. 4 -

+

'b: '., :., ::,. -..

-.,,, ~

~ ^ ' ~

~

SPECIAL TEST EXCEPTIONS 3/4.10.3 REACTOR COOLANT LOOPS i

LIMITING CONDITION FOR OPERATION 3.10.3 T!ie limitations of Specification 3.4.1.1 and the.noted requirements of l

Table 2.2-1 and Table 3.3-1 may be suspended during the performance of startup and PHYSICS TESTS, provided:

s a.

The THERMAL POWER does not exceed 5% of RATED THERMAL POWER, and b.

The reactor trip setpoints of the OPERABLE power level channels are set at less than or equal to 20% of RATED THERMAL POWER.

APPLICABILITY:

During startup and PHYSICS TESTS.

ACTION:

With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately trip the reactor.

SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5%

of RATED THERMAL POWER at least once per hour during startup and PHYSICS TESTS.

4.10.3.2 Each logarithmic and linear power level neutron flux monitoring.

channel shall be' subjected to a CHANNEL FUNCTIONAL TEST within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating startup and PHYSICS TESTS.

l l

SAN ONOFRE-UNIT 2 3/4 10-3 Amendment No. 4

SPECIAL TEST EXCEPTIONS 3/4.10.4 CENTER CEA MISALIGNMENT I

i LIMITING CONDITION FOR OPERATION 3.10.4 The requirements of Specifications 3.1.3.1 and 3.1.3.6 may be suspended during the performance of PHYSICS TESTS to determine the isothermal temperature coefficient, moderator temperature coefficient and power coefficient provided:

a.

Only the center CEA (CEA #1) is misaligned, and b.

The limits of Specification 3.2.1 are maintained and' determined as specified in Specification 4.10.4.2 below.

APPLICA8ILITY: MODES I and 2.

ACTION:

With any of the limits of Specification 3.2.1 being exceeded while the requirements of Specifications 3.1.3.1 and 3.1.3.6 are suspended, either:

a.

Reduce THERMAL POWER sufficiently to satisfy the requirements of Specification 3.2.1, or b.

Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS L

4.10.4.1 The THERMAL POWER shall be determined at least once per hour during PHYSICS TESTS in which the requirements of Specifications 3.1.3.1.and/or 3.1.3.6 are suspended and shall be verified to be within the test power plateau.

4.10.4.2 The linear heat rate shall be determined to be within the limits of Specification 3.2.1 by monitoring it continuously with the Incore Detector Monitoring System pursuant to the requirements of Specification 3.3.3.2 during PHYSICS TESTS above 5% of RATED THERMAL POWER in which the requirements of Specifications 3.1.3.1 and/or 3.1.3.6 are suspended.

i i

SAN ONOFRE-UNIT 2-3/4 10-4 i

.:~---

~. : --

o SPECIAL TEST EXCEPTIONS 3/4.10.5 RADIATION MONITORING / SAMPLING LIMITING CONDITION FOR OPERATION 3.10.5 The OPERABILITY requirements of Specifications 3/4.3.2, 3/4.3.3.1, 3/4.3.3.6, 3/4.3.3.8, and 3/4.3.3.9 for the radiation monitoring and sampling instrumentation listed in Table 3.10-1 may be modified per Table 3.10-1 provided the requirements listed in Table 3.10-1 are met.

1 APPLICABILITY:

As shown in Table 3.10-1.

l ACTION:

With the THERMAL POWER or criticality condition exceeding the limit for monitoring / sampling instrumentation as shown in Table 3.10-1, immediately trip the reactor.

SURVEILLANCE REQUIREMENTS 4.10.5 The monitoring / sampling instrumentation listed in Table 3.10-1 shall be demonstrated OPERABLE accordance with Specification 4.3.2, 4.3.3.1, 4.3.3.6,-

4.3.3.8 or 4.3.3.9, as applicable, except as modified by Table 3.10-1.

3 I

i j

SAN ONOFRE-UNIT 2 3/4 10-5 l

TABLE 3.10-1 RADIATION MONITORING / SAMPLING EXCEPTIONS 1.

Testing performed pursuant to FSAR Section 11.5.2.1.5.2 in startup program shall satisfy the initial CHANNEL CALIBRATION for the following monitors prior to first exceeding 5% RATED THERMAL POWER:

a.

Control Room Airborne Monitors 2RT-7824 2RT-7825 b.

Containment Airborne Monitors 2RT-7804-1 2RT-7807-2 c.

Containment Purge Area Monitors 2RT-7856-1 2RT-7857-2 d.

Containment Area Radiation -

2RT-7820-1 High Range Monitors 2RT-7820-2 e.

Plant Vent Stack Airborne Monitor 2/3RT-7808 f.

Radwaste Discharge Line Monitor 2/3RT-7813 g.

Blowdown Neutralization Sump Monitor 2R T-7817 h.

Turbine Building Sump Monitor 2RT-7821 2.

The following monitors and samplers shall be OPERABLE prior to first exceeding 5% RATED THERMAL POWER:

a.

Main Steam Line Area Monitors 2RT-7874A1 lj 2RT-7874B1 2RT-7875A1 2RT-7875B1 b.

Condenser Evacuation System -

2RT-7870-1 Wide Range Monitor c.

Purge / Vent Stack Monitors -

2RT-7865-1 Wide Range 3RT-7865-1 d.

Plant Vent Stack Flow Rate Monitor e.

_ Containment Purge Flow Rate Monitor f.

Condenser Evacuation System Iodine Sampler Particulate Sampler Flow Rate Monitor 3.

T'he Steam Jet Air Ejector Monitor (2RT-7818) shall be OPERABLE prior to initial criticality.

SAN ON0FRE-UNIT 2 3/4 10-6 Amendment No. 4 j

TABLE 3.;0-1 (Continued) 4.

Testing performed pursuant to FSAR Sectino 14.2.12 in startup program is acceptable for the initial CHANNEL FUNCTIONAL TEST for a period up to 30 days following initial criticality for the following liquid effluent monitors:

a.

Radwaste Discharge Line Monitor 2/3 RT-7813 b.

Blowdown Neutralization Sump Monitor 2RT-7817 c.

Turbine Building Sump Monitor 2RT-7821 5.

Continuous monitoring and sampling of the containment purge exhaust directly from the purge. stack shall be provided for the low and high volume (8-inch and 42-inch) containment purge prior to startup following the first refueling outage.

Containment airborne monitor 2RT-7804-1 or l

2RT-7807-2 and associated sampling media shall perform these functions prior to initial criticality.

From initial criticality to the startup following the first refueling outage containment airborne monitor 2RT-7804-1 and associated sampling media shall perform the above required functions.

SAN ON0FRE-UNIT 2 3/4 10-7 Amendment No. 4

SPECIAL TEST EXCEPTIONS MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.10.6 The minimum temperature for criticality limits of Specification 3.1.1.4 l

and the MODE 2 definition of Table 1.1 may be suspended during low temperature PHYSICS TESTS to a minimum temperature of 320 F provided:

a.

The THERMAL POWER does not exceed 5% of RATED THERMAL POWER.

b.

The reactor trip setpoints on the OPERABLE Linear Power Level - High neutron flux monitoring channels are set at 5 20% of RATED THERMAL POWER, and The Reactor Coolant System temperature and pressure relationship and c.

the minimum temperature for criticality are maintained within the acceptable region of operation shown on Figure 3.4-2.

APPLICABILITY: MODE 2.*

ACTION:

With the THERMAL POWER > 5 percent of RATED THERMAL POWER, immediately a.

trip the reactor, b.

With the Reactor Coolant System temperature and pressure relationship and/or the minimum temperature for criticality within the region of unacceptable operation on Figure 3.4-2, immediately trip the reactor and, if necessary, restore the temperature pressure relationship to within its limit within 30 minutes; perform the engineering evaluation required by Specification 3.4.8.1 prior to the next reactor criticality.

g SURVEILLANCE REQUIREMENTS 4.10.6.1 At least once per hour:

The Reactor Coolant System temperature and pressure relationship and a.

the minimum temperature for criticality shall be verified to be within the acceptable region for operation of Figure 3.4-2.

b.

The THERMAL POWER shall be determined to be 1 5% of RATED THERMAL POWER.

c.

The Reactor Coolant System temperature shall be verified to be greater than or equal to 320 F.

4.10.6.2 Each Logarithmic Power Level and Linear Power Level channel shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating low temperature PHYSICS TESTS.

"First c. ore only, prior to first exceeding 5% RATED THERMAL POWER.

SAN ON0FRE-UNIT 2 3/4 10-8 Amendment Nn. 4

TABLE 3.12-1 (Continued)

!S 2:

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM E

R?

Number of Samples E

Exposure Pathway

~ and Sampling and a

a 2{

and/or Sample Sample Locations Collection Frequency Type and Frequency of Analyses o'

4 INGESTION a.

Nonmigratory 3 Locations One sample in season, or at Gamma isotopic analysis on Marine least once per 184 days if not edible portions.

Animals seasonal.

One sample of each of the following species:

1. Fish-2 adult species such as perch or sheepshead.

2. Crustaceae-such as crab or lobster.

3. Mollusks-such as limpets or t'

seahares,

,a R;

b.

Local Crops 2 Locations Representative vegetables, Gamma isotopic analysis on J,

normally 1 leafy and I fleshy edible portions semiannually collected at harvest time.

At and I-131 analysis for leafy least 2 vegetables collected crops.

semiannually from each location.

5I

-3 i

en O

-=

. TABLE 3.12-1 (Continued)

TABLE NOTATION O

i S$

q$

a.

Sample locations are indicated in the ODCM C5 b.

Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.

ro c.

The purpose of this sample is to obtain background information.

If it is not practical to establish control locations in accordance with the distance and wind direction criteria, other sites which pro-vide valid background data may be substituted.

d.

Canisters for the collection of radioiodine in air are subject to channeling.

These devices should be carefully checked before operation in the field or several should be mounted in series to prevent loss of iodine.

e.

Regulatory Guide 4.13 provides minimum acceptable performance criteria for thermoluminesc?nce dosim-ki etry (TLD) systems used for environmental monitoring.

One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addi-U tion to, integrating dosimeters.

For the purposed of this table, a tharmoluminescent dosimeter may a

be considared to be one phosphor and two or more phosphors in a packet may be considered as two or more dosimeters.

Film badges saould not be used for measuring direct radiation.

f.

Composite samples should be collected with equipment (or equivalent) which is capable of collecting an aliquot at time intervals which are very short (e.g., hourly) relative to the compositing period (e.g., monthly).

2r a

8r

.O

PLANT SYSTEMS BASES FIRE SUPPRESSION SYSTEMS (Continued)

In the event the fire suppression water system becomes inoperable, immediate corrective measures must be taken since this system provides the major fire suppression capability of the plant.

The requirement for a twenty-four hour report to the Commission provides for prompt evaluation of the acceptability of the corrective measures to provide adequate fire suppression capability for the continued protection of the nuclear plant.

The San Onofre Unit 2&3 fire pumps and water supplies, supply water to the San Onofre Unit 1 fire system.

3/4.7.9 FIRE RATED ASSEMBLIES The OPERABILITY of the fire barriers and barrier penetrations ensure that fire damage will be limited.

These design features minimize the possibility of a single fire involving more than one fire area prior to detection and extinguishment.

The fire barriers, fire barrier penetrations for conduits, cable trays and piping, fire windows, fire dampers, and fire doors are periodically inspected to verify their OPERABILITY.

i SAN ON0FRE-UNIT 2 8 3/4 7-7 Amendment No. 4

3/4.8 ELECTRICAL POWER SYSTEMS BASES 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES and ONSITE POWER DISTRIBUTION SYSTEMS The OPERABILITY of the A.C. and D.C. power sources and associated distri-bution systems during operation ensures that sufficient power will be available to supply the safety related equipment required for 1) the safe shutdown of the facility and 2) the mitigation and control of accident conditions within the facility.

The minimum specified independent and redundant A.C. and D.C.

power sources and distribution systems satisfy the requirements of General Design Criteria 17 of Appendix "A" to 10 CFR 50.

The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commen-surate with the level of degradation.

The OPERABILITY of the power sources are consistent with the initial condition assumptions of the safety analyses and are based upon maintaining at least one redundant set of onsite A.C. and D.C. power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss of offsite power and single failure of the other onsite A.C. source.

The A.C. and D.C. source allowable out-of-service times are based on Regulatory Guide 1.93, " Availability of Electrical Power Sources," December 1974. When'one diesel generator is inoperable, there is an additional ACTION requirement to verify that all required systems, subsystems, trains, components and devices, that depend on the remaining OPERABLE diesel generator as a source of emergency power, are also OPERABLE, and that the steam-driven auxiliary feedwater pump is OPERABLE.

This require-ment is intended to provide assurance that a loss of offsite power event will not result in a complete loss of safety function of critical systems during the period one of the diesel generators is inoperable.

The term verify as used in this context means to administratively check by examining logs or other information to determine if certain components are out-of-service for maintenance or other reasons.

It does not mean to perform the surveillance requirements needed to demonstrate the OPERABILITY of the component.

The OPERABILITY of the minimum specified A.C. and D.C. power sources and associated distribution sy, stems during shutdown and refueling ensures that

1) the facility can be maintained in the shutdown or refueling condition for i

extended time periods and 2) sufficient instrumentation and control capability l

is available for monitoring and maintaining the unit status.

The Surveillance Requirements to verify OPERABILITY of the required independent circuits between the offsite transmission network and the onsite Class 1E distribution system allows for one of two alternatives.

The connection can be made by back-feeding from Unit 3.

Alternatively, the Unit 2 auxiliary transformer also may provide an alternate means of operation during low power PHYSICS TESTS. With the Unit 2 isolated phase bus links removed, if preferred power from the Unit 2 reserve auxiliary transformer is lost, the 4.16 kV feeder circuit breaker can be inserted into the auxiliary transformer position

.to reestablish power to the Class IE bus.

Breaker controls for this connection, i

SAN ON0FRE-UNIT 2 B 3/4 8-1 I

i L __ ; ~ ---- - _

~ ~.. -..

ELECTRIC POWER SYSTEMS BASES A.C. SOURCES, D.C. SOURCES AND ONSITE POWER DISTRIBUTION SYSTEMS (Continued) as well as operation of loss of voltage logic, is the same as for the primary connection using the reserve auxiliary transformer, with the exception of no transfer to the companion unit.

The Surveillance Requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of Regulatory Guides 1.9 " Selection of Diesel Generator Set Capacity for Standby Power Supplies," March 10, 1971, and 1.108 " Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants," Revision 1, August 1977, and 1.137, " Fuel Oil Systems for Standby Diesel Generators,"

Revision 1, October 1979.

Reg. Guide 1.137 recommends testing of fuel oil samples in accordance with ASTM-0270-1975. However, ASTM-0270-1965 was reverified in 1975 rather than re-issued.

The reverified 1965 standard is therefore the appro-priate standard to be used.

Additionally, Regulatory Guide 1.9 allows loading of the diesel generator to its 2000 hour0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months rating in an accident situation.

The full load, continuous operation rating for each diesel generator is 4700 kw, while the calculated acci-dent loading is 4000 kw.

No 2000 hour0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months loading has been specified by 'the diesel generator manufacturer and, as a result the full loading rating of 4700 kw is conservatively established as the 2000 hour0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months rating.

Diesel frequency droop restrictions are established due to HPSI flow rate considerations.

The Surveillance Requirement for demonstrating the OPERABILITY of the Station batteries are based on the recommendations of Regulatory Guide 1.129,

" Maintenance Testing and Replacement of Large Lead Storage Batteries for Nuclear Power Plants," February 1978, and IEEE Std 450-1980, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations."

Verifying average electrolyte temperature above the minimum for which the battery was sized, total battery terminal voltage onfloat charge, connection resistance values and the performance of battery service and discharge tests ensures the effectiveness of the charging system, the ability to handle high discharge rates and compares the battery capacity at that time with the rated capacity.

Table 4.8-2 specifies the normal limits for each designated pilot cell and each connected cell for electrolyte level, float voltage and specific gravity.

The limits for the designated pilot cells float voltage and specific gravity, greater than 2.13 volts and.015 below the manufacturer's full charge specific gravity or a battery charger current that had stabilized at a low value, is characteristic of a charged cell with adequate capacity.

The normal limits for each connected cell for float voltage and specific gravity, greater than 2.13 volts and not more than.020 below the manufacturer's full charge specific gravity with an average specific gravity of all the connected cells not more than.010 below the manufacturer's full charge specific gravity, ensures the OPERABILITY and capability of the battery.

SAN ONOFRE-UNIT 2 B 3/4 8-2 Amendment No. 4

3/4.10 SPECIAL TEST EXCEPTIONS BASES 3/4.10.1 SHUTDOWN MARGIN This special test exception provides that a minimum amount of CEA worth is immediately available for reactivity control when tests are performed for CEAs worth measurement.

This special test exception is required to permit the periodic verification of the actual versus predicted core reactivity condition occurring as a result of fuel burnup or fuel cycling operations.

3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS This special test exception permits individual CEAs to be positioned outside of their normal group heights and insertion limits during the performance of such PHYSICS TESTS as those required to 1) measure CEA worth and 2) determine the reactor stability index and damping factor under xenon oscillation conditions.

3/4.10.3 REACTOR COOLANT LOOPS This special test exception permits reactor criticality under no flow conditions and is required to perform certain startup and PHYSICS TESTS while at low THERMAL POWER levels.

3/4.10.4 CENTER CEA MISALIGNMENT This special test exception permits the center CEA to be misaligned during PHYSICS TESTS required to determine the isothermal temperature coefficient and power coefficient.

3/4.10.5 RADIATION MONITORING / SAMPLING This special test exception permits fuel loading and reactor operation with radiation monitoring / sampling instrumentation calibration and quality assurance conforming to either FSAR procedures or Regulatory Guide 4.15 Rev 1, February 1979.

This test exception is required to a,llow for a phased implementa-tion of Regulatory Guide 4.15 Rev.1, February 1979.

Equivalent instrumentation, quality assurance and/or calibration is provided until full implementation of Regulatory Guide 4.15 Rev.1, February 1979.

The containment airb.orne monitors and associated sampling media test exception is required to allow for operation prior to and during installation of upgraded monitors / media.

Adequate monitoring is provided until and subsequent to the completion of the upgraded installation.

Extensive containment air mixing during high volume purge (MODES 5 and 6) occurs as a result of containment HVAC and fans resulting in representative air monitoring via either 2RT-7804-1 or 2RT-7807-2.

During low volume purge operations (MODES 1, 2, 3 and 4) 2RT-7804-1 provides representative indication of purged air due to its location in the immediate vicinity of the low volume purge exhaust.

SAN ON0FRE-UNIT 2 8 3/4 10-1 Amendment No. 4

3/4.10 SPECIAL TEST EXCEPTIONS BASES 3/4.10.6 MINIMUM TEMPERATURE FOR CRITICALITY This special test exception permits reactor criticality at low THERMAL POWER levels with T' verify the adequacy of design codes for new fuel des below 520 F during PHYSICS-TESTS which provide data that can be used to for reduced temperature conditions.

The Low Power Physics Testing program at low temperature (320 F) is used to perform the following tests:

1.

Biological shielding survey test 2.

Isothermal temperature coefficient tests 3.

Regulatory CEA group tests 4.

Boron worth tests 5.

Critical configuration boron concentration SAN ON0FRE-UNIT 2 8 3/4 10-2

ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY i

6.1.1 The Station Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

6.1. 2 The Shift Supervisor (or during his absence from the Control Room Area, l

a designated individual) shall be responsible for the Control Room command function.

A management directive to this effect, signed by the Vice-President of Nuclear Operations shall be reissued to all station personnel on an annual 1

basis.

6.2 ORGANIZATION 0FFSITE i

6.2.1 The offsite organization for unit management and technical support shall be as shown in Figure 6.2-1.

UNIT STAFF 6.2.2 The Unit organization shall be as shown in Figure 6.2-2 and:

a.

Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.

b.

At least one licensed Reactor Operator shall be in the Control Room when fuel is in the reactor.

In addition, while the unit is in MODE 1, 2, 3 or 4, at least one licensed Senior Reactor Operator shall be in the Control Room area identified as such on Table 6.2-1.

A health physics technician # shall be on site when fuel is in the b

c.

reactor.

d.

All CORE ALTERATIONS shall be observed and,directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

A site Fire Bpigade of at least 5 members shall be maintained onsite e.

at all times.

The Fire Brigade shall not include the Shift Super-I visor and the 2 other members of the minimum shift crew necessary for l

safe shutdown of the unit and any personnel required for other essential functions during a fire emergency.

The health physics technician and Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence provided immediate action is taken to fill the required positions.

SAN ONOFRE-UNIT 2 6-1 Amendment No. 4

~.-

CHAIRMAN OF THE BOARD I

SENIGH VICE PRESIDENT I

I I

I VICE PRESIDENT VICE PRESIDENT VICE PRESIDENT VICE PRESIDENT VICE PRESIDENT VICE PRESIDENT (FUEL SUPPLY)

(NUCLEAR ENGINEERING ISYSTEM (ENGINEERING &

(ADVANCED (POWER SUPPLY)

AND OPER ATIONS)

DEVELOPMENT)

CONTRUCTION)

ENGINEERING)

I MANAGER OF MANAGER OF MANAGER OF MANAGER OF NUCLEAR ENGINEERING, NUCLEAR ENVIRONMENTAL ENGINEERING SAFETY & LICENSING OPERATK NS AFFAIRS DESIGN DIRECTOR OF

MANAGER, TRAINING RESEARCH AND QUALITY MANAGER DEVELOPMENT ASSURANCE LEAR MANAGER PROJECT M AN AGE R

MANAGER, NUCLEAR LICEbSING SAN ONOFRE UNITS 2 & 3 AFETY SY TEMS R & D I

HEADOUARTERS NUCLEAR SAFETY STAFF GROUP

L---

I STATION ON SITE STAFF REVIEW A GER COMMITTEE Figure 6.21 OFFSITE ORGANIZATION SAN ONOFRE NUCLEAR GENER ATING STATION - UNIT 2 SAN ONOFRE-UNIT 2 6-2 l

STATION MANAGER DEPUTY STATION l

MANAGER I

I I

MANAGER.

M AN AGER. 4 M AN AGE R.

MATERIAL AND CONFIGURATION STATION STATION ADMINISTRATIVE CONTROL AND EMERGENCY SECURITY SERVICES COMPLIANCE PREPAREDNESS I

I I

M AN AGER.

M AN AGE R.

MANAGER.

HEALTH OPERATIONS MAINTENANCE TECHNICAL pH PLANT 1

SUPERVISING COMPUTER SUPT.

SONGS 263 ENGINEER yp SUPERVISOR

'==

MAINTENANCE PLANNING SUPERVISBNG SUPERVISOR ENGINEER NSSS RADWASTE SUPERVISOR SUPERVISING SUPERVISOR OF ENGINEER NUCLEAR PLANT NSSS SUPPORT DOSIMETRY MAINTENANCE i,

SONGS 263 SUPERVISING ENGINEER 2

I SUPERVISOR SHIFT TECH.

SHIFT ADVISORS H.Pc ENGR'S OF PLANT SUPERVISORS GROUP

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COORDINATION SUPERVISOR g

SUPERVISOR 2

OP "^

COORDINATORS PORE N

I SUPERVISOR OF 3

$H CHEMISTRY PER$

1. AT TIME OF APPOINTMENT TO THE POSITION. SENIOR REACTOR OPERATOR LICENSE REQUIRED.

2. SENIOR REACTOR LICENSE REQUIRED.

3 CONTROL AND ASSISTANT CONTROL OPERATORS ARE HOLDERS OF REACTOR OPERATOR LICENSES.

4 INCLUDES FIRE PROTECTION.

Figure 6.2-2 Unit Organization, San Onofre Nuclear Generting Station Unit 2 SAN ONOFRE-UNIT 2 6-3 Amendment f!O. 4

Table 6.2-1 MINIMUM SHIFT CREW COMPOSITION POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODES 1, 2, 3 & 4 MODES 5 & 6 SS 1

1 SRO 1

None R0 2

1 A0 2

1 STA 1

None SS Shift Supervisor with a Senior Reactor Operators License on l

Unit 2 SRO -

Individual with a Senior Reactor Operators License on Unit 2 R0 Individual with a Reactor Operators License on Unit 2 A0 Auxiliary Operator STA Shift Technical Advisor Except for the Shift Supervisor, the Shift Crew Composition may be one less l

than the minimum requirements of Table 6.2-1 for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the Shift Crew Composi-tion to within the minimum requirements of Table 6.2-1.

This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.

During any absence of the Shift Supervisor from the Control Room Area shown in l

Figure 6.2-3 while the unit is in MODE-1, 2, 3 or 4, an individual (other than the Shift Technical Advisor) with a valid SRO license shall be designated to assume the Centrol Room command function.

During any absence-of the Shift Supervisor from the Control Room Area shown in Figure 6.2-3 while the unit is in MODE 5 or 6, an individual with a valid SRO or RO license shall be designated to assume the Control Room command function.

Amendment No. 4 SAN ON0FRE-UNIT 2 6-4

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ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG) i i

FUNCTION 6.2.3.1 The ISEG shall function to examine plant operating characteristics, NRC issuances, industry advisories, Licensee Event Reports and other sources j

of plant design and operating experience information which may indicate areas for improving plant safety.

COMPOSITION 6.2.3.2 The ISEG shall be composed of at least five dedicated full-time engineers.

Each shall have a Bachelor's Degree in Engineering or Physical Science and at least two years professional level experience in his field.

Off-duty qualified Shift Technical Advisors may be used to fulfill this requirement.

RESPONSIBILITIES 6.2.3.3 The ISEG shall be responsible for maintaining surveillance of plant activities to provide independent verification

that these activities are performed correctly and that human errors are reduced as much as practical.

AUTHORITY 6.2.3.4 The ISEG shall make detailed recommendations for revised procedures, equipment modifications, maintenance activities, operations activities or other means of improving plant safety to the Supervisor, Nuclear Safety Group.

1 6.2.4 SHIFT TECHNICAL ADVISOR I

(

The Shift Technical Advisor shall provide technical support to the Shift l

Supervisor in the areas of thermal hydraulics, reactor engineering and plant i

analysis with regard to the safe operation of the unit.

The Shift Technical i

Advisor shall have a Bachelor's Degree or equivalent in a scientific or engineering discipline with specific training in plant design and in the response and analysis of the plant for transients and accidents.

6.3 UNIT STAFF QUALIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for the Health Physics Manager who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975.

i A

Not responsible for sign-off function.

SAN ONOFRE-UNIT 2 6-5 Amendment"No. 4

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ADMINISTRATIVE CONTROLS 6.4 TRAINING 6.4.1 A retraining and replacement training program for the unit staff shall be maintained under the direction of the Manager, Nuclear Training and shall meet or exceed the requirements and recommendations of Sections 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55 and the supplemental requirements specified in Section A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant industry operational experience identified by the ISEG.

6.5 REVIEW AND AUDIT 6.5.1 ONSITE REVIEW COMMITTEE (OSRC)

FUNCTION 6.5.1.1 The Onsite Review Committee shall function to advise the Station Manager on all matters related to nuclear safety.

COMPOSITION 6.5.1.2 The Onsite Review Committee shall be composed of the:

Chairman:

Station Manager Member:

Deputy Station Manager Member:

Manager, Operations Member:

Manager, Technical Member:

Plant Superintendent SONGS Unit 2 & 3 Member:

Supervisor of I&C Member:

Manager, Health Physics l

Member:

Supervisor of Chemistry Member:

Manager, Maintenance l

Member:

Supervising Engineer (NSSS, NS35 Support, Computer, or STA)

Member:

SanDiegoGas&{}ectricRepresentative, Senior Engineer ALTERNATES

6. 5.1. 3 All alternate members shall be appointed in writing by the OSRC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting memDers in OSRC activities at any one time.

(1)BS degree in Engineering or Physical Science plus at least four years professional level experience in his field.

At least one of the tour years experience shall be nuclear power plant experience.

SAN ONOFRE-UNIT 2 6-6 Amendment No. 4

.=

ADMINISTRATIVE CONTROLS j

MEETING FREQUENCY 6.5.1.4 The OSRC shall meet at least once per calendar month and as convened by the OSRC Chairman or his designated alternate.

QUORUM 6.5.1.5 The minimum quorum of the OSRC necessary for the performance of the OSRC responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or his designated alternate and four members including alternates.

RESPONSIBILITIES

6. 5.1. 6 The Onsite Review Committee shall be responsible for:

i a.

Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Nuclear Safety Group (NSG).

b.

Review of events requiring 24-hour written notification to the Commission.

Review of unit operations to detect potential nuclear safety hazards.

c.

d.

Performance of special reviews, investigations or analyses and reports thereon as requested by the Station Managar or the NSG.

Review and documentation of judgment concerning. prolonged operation in e.

bypass, channel trip, and/or repair of defective protection channels of process variables placed in bypass since the last OSRC meeting.

f.

Review and approval of using and entering values of CPC addressable constants outside the allowable range of Table 2.2-2.

i l

l SAN ONOFRE-UNIT 2 6-7

ADi4INIST.RATIVE CONTROLS AUTHORITY 6.5.1.7 The Onsite Review Committee (OSRC) shall:

Render determinations in writing with regard to whether or not items i

a.

considered under 6.5.1.6(a) above constitute unreviewed safety questions.

b.

Provide written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the Manager of Nuclear Operations and NSG of disagreement between the OSRC and the. Station Manager; however, the Station Manager shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.

RECORDS 6.5.1.8 The Onsite Review Committee shall maintain written minutes of each OSRC meeting that, at a minimum, document the results of all OSRC activities performed under the responsibility and authority provisions of these technical specifications.

Copies shall be provided to the Nuclear Safety Group.

6.5.2 TECHNICAL REVIEW AND CONTROL ACTIVITIES 6.5.2.1 The Station Manager shall assure that each procedure and program required by Specification 6.8 and other procedures which affect nuclear safety, and changes thereto, is prepared by a qualified individual / organization.

Each such procedure, and changes thereto, shall be reviewed by an individual / group other than the individual / group which prepared the procedure, or changes thereto, but who may be from the same organization as the individual / group which prepared the procedure, or changes thereto.

6.5.2.2 Proposed changes to the Appendix "A" Technical Specifications shall be prepared by a qualified individual / organization.

The preparation of each proposed Technical Specifications change shall be reviewed by an individual /

group other than the individual / group which prepared the proposed change, but who may be from the same organization as the individual / group which prepared the proposed change.

Proposed changes to the Technical Specifications shall be approved by the Station Manager.

6.5.2.3 Proposed modifications to unit nuclear safety-related structures, systems and components shall be designed by a qualified individual /

organization.

Each such modification shall be reviewed by an individual / group other than the individual / group which designed the modification, but who may be from the same organization as the individual / group which designed the modifi-cation.

Proposed modifications to nuclear safety-related structures, systems and components shall be approved prior to implementation by the Station Manager; or by the Manager, Technical as previously designated by the Station Manager.

A SAN ON0FRE-UNIT 2 6-8 Amendment f!o. 4 L

ADMINISTRATIVE CONTROLS ACTIVITIES (Continued) 2 6.5.2.4 Individuals responsible for reviews performed in accordance with 6.5.2.1, 6.5.2.2 and 6.5.2.3 shall be members of the station supervisory staff, previously designated by the Station Manager to perform such reviews.

Each such review shall include a determination of whether or not additional, cross-disciplinary, review is necessary.

If deemed necessary, such review shall be performed by the appropriate designated review personnel.

6.5.2.5 Proposed 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, the Manager, Technical, the Manager, Operations, the Manager, Maintenance, the Deputy Station Manager or the Manager, Health Physics as previously designated by the Station Manager.

6.5.2.6 The station security program, and implementing procedures, shall be reviewed at least once per 12 months.

Recommended changes shall be approved by the Station Manager and transmitted to the Manager of Nuclear Operations and to the NSG.

i 6.5.2.7 The station emergency plan, and implementing procedures, shall be reviewed at least once per 12 months.

Recommended changes shall be approved by the Station Manager and transmitted to the Manager of Nuclear Operations and to the NSG.

6.5.2.8 The Station Manager shall assure the performance of a review by a qualified individual / organization of every unplanned onsite release of radio-active material to the environs including the preparation and forwarding of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence to the Manager of Nuclear Operations and to the t

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

I 6.5.2.10 Reports documenting each of the activities performed under Specifications 6.5.2.1 through 6.5.2.9 shall be maintained.

Copies shall be provided to the Managec of Nuclear Operations and the Nuclear Safety Group.

6.5.3 NUCLEAR SAFETY GROUP (NSG)

FUNCTION 6.5.3.1 The Nuclear Safety Group shall function-to provide independent review j

and audit of designated activities in the areas of:

1 SAN ONOFRE-UNIT 2 6-9 Amendment No. 4 l

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ADMINISTRATIVE CONTROLS FUNCTION (Continued)'

a.

nuclear power plant operations b.

nuclear engineering c.

chemistry and radiochemistry d.

metallurgy e.

instrumentation and control f.

radiological safety g.

mechanical and electrical engineering h.

quality assurance practices COMPOSITION 6.5.3.2 NSG shall consist of a Supervisor and at least tnree staff specialists.

The Supervisor shall have a Bachelor's Degree in Engineering or Physical Science and a minimum of six years of professional level mdnagerial experience in the power field.

Each staff specialist shall have a Bachelor's Degree in Engineer-ing or Physical Science and a minimum of five years,of professional level experience in the field of his specialty.

The NSG shall use specialists from other technical organizations to augment its expertise in the disciplines of 6.5.2.1.

Such specialists shall meet the same qualification requirements as the NSG members.

CONSULTANTS 6.5.3.3 Consultants shall be utilized as determined by the NSG Supervisor to provide expert advice to the NSG.

REVIEW 6.5.3.4 The NSG shall review:

The safety evaluations for 1) changes to procedures required by a.

Specification 6.8, equipment or systems and 2) tests or experiments completed under the provision of Section 50.59, 10 CFR_, to verify that such actions did not constitute an unreviewed safety question.

b.

Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.

Proposed tests or experiments which involve an unreviewed safety c.

question as defined in Section 50.59, 10 CFR.

SAN ONOFRE-UNIT 2 6-10 Amendment No. 4

ADMINISTRATIVE CONTROLS 6.6 REPORTABLE OCCURRENCE ACTION 6.6.1 The following actions shall be taken for REPORTABLE OCCURRENCES:

The Commission shall be notified and/or a report submitted pursuant a.

to the requirements of Specification 6.9.

b.

Each REPORTABLE OCCURRENCE requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months notification to the Commission shall be reviewed by the OSRC and submitted to the NSG and the Manager of Nuclear Operations.

6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

The NRC Operations Center shall be notified by telephone as soon as a.

possible and in all cases within one hour.

The Manager of Nuclear Operations and the NSG Chairman shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

A Safety Limit Violation Report shall be prepared.

The report shall be reviewed by the OSRC.

This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

c.

The. Safety Limit Violation Report shall be submitted to the Commission, the Manager of Nuclear Operations and the NSG within 14 days of the vioTation.

d.

Critical operation of the unit shall not be resumed until authorized F.

by the Commission.

6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

The applicable procedures recommended in Appendix "A" of Regulatory a.

-Guide 1.33, Revision 2, February 1978.

b.

Refueling operations.

Surveillance and test activities of safety related equipment.

c.

4 d.

Security Plan implementation.

e.

Emergency Plan implementation.

f.

Fire Protection Program implementation.

SAN ONOFRE-UNIT 2 6-13

-: ~

~

ADMINISTRATIVE CONTROLS g.

PROCESS CONTROL PROGRAM implementation.*

h.

OFFSITE DOSE CALCULATION MANUAL implementation.

i.

Quality Assurance Program for effluent and environmental monitoring, s

using the guidance in Regulatory Guide 4.15 Rev. 1, February 1979.

NOTE:

Quality Assurance Program for effluent and environmental monitoring and sampling shall be in accordance with Regulatory Guide 4.15, December, 1977 prior to first exceeding 5% RATED THERMAL POWER or July 1,1982, whichever occurs first; subsequent to this time the Quality Assurance Program shall be in accordance with Regulatory Guide 4.15, Rev. 1, February,1979.

J.

Modification of Core Protection Calculator (CPC) Addressable Constants.

NOTE:

Modification to the CPC addressable constants based on information obtained through the Plant Computer - CPC data link shall not be made without prior approval of the Onsite Review Committee.

6.8.2 Each procedure of 6.8.1 above, and changes thereto, shall be approved by the Station Manager; or by (1) the. Manager, Operations (2) the Manager, Technical (3) the Manager, Maintenance, (4) the Deputy Station Manager, or (5) the Manager, Health Physics as previously designated by the Station Manager; prior to implementation and shall be reviewed periodically as set forth in administrative procedures.

6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:

a.

The intent of the original procedure is not altered.

b.

The change is approved by two members of the plant maaagement staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.

The change is documented, reviewed and approved by the Station c.

Manager; or by (1) the Deputy Station Manager, (2) the Manager, Operations, (3) the Manager, Maintenance, (4) the Manager, Technical, or (5) the Manager, Health Physics as previously designated by the Station Manager; within 14 days of implementation.

6.8.4 The following programs shall be established, implemented, and maintained:

a.

Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels.

The systems include the high pressure safety injection recirculation, the shutdown cooling system, the reactor coolant sampling system (post-accident sampling piping only), the containment spray system, the radioactive waste gas system (post-accident sampling return piping only) and the liquid radwaste system (post accident sampling return piping only).

The program shall include the following:

(i) Preventive maintenance and periodic visual inspection requirements, and (ii) Integrated leak test requirements for each system at refueling cycle intervals or less.

See Specification 6.13.1 SAN ONOFRE-UNIT 2 6-14 Amendment No. 4

ADMINISTRATIVE CONTROLS h.

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

i.

Records of Quality Assurance activities required by the QA Manual.

j.

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

k.

Records of meetings of the OSRC and the NSG.

1.

Records of the service lives of all snubbers listed in Tables 3.7-4a and 3.7-4b including the date at which the service life commences and associated installation and maintenance records.

m.

Recort of secondary water sampling and water quality.

6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

6.12 HIGH RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a high radiation area and i

entrance thereto shall be controlled by requiring issuance of a Radiation Exposure' Permit (REP)*

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

A radiation monitoring device which continuously indicates the a.

radiation dose rate in the area.

l b.

A radiation monitoring device which continuously integrates the l

radiation dose rate in the area and alarms when a preset integrated dose is received.

Entry into such areas with this monitoring aevice may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.

^ Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt from the REP issuance requirement during the performance l

of their assigned radiation protection duties, provided they are otherwise following aporoved plant radiation protection procedures for entry into high i

radiation areas.

SAN ONOFRE-UNIT 2 6-23 L

s E'

N'

.\\,

7 N

ADMINISTRATIVE CONTROLS is AnindividualqualifiNdinradiationprotectionprocedureswhois sy, c

c.

equipped with a radiation dose rate monitoring device who is responsible for providing positive control over the activities within the area ano shall perform periodic radiation surveillance ~'at the frequency specified by the facility Health Physicist in the Radiation Exposure Permit.

t 6.12.2 In addition to the requirements.of 6.12.1, areas accessible,to personnel with radiation levels such.that a major portion of the body could receive in one hour a dose greater than 1000 mrem shall be provided with. locked doors to prevent unauthorized entry, and the keys shall be maintained undes the administrative control of the Shift Supervisor on duty and/or healthlph'ysics-supervision.

Doors shall remain locked except during periods of access by personnel under an l

approved REP which shall specify the dose rate levels in the immediate work area and the maximum allowable stay time for individuals in that area.

t For individual areas accessible to personnel with radiation levels.such that a major portion of the body could receive in one hour a dose in excess of 1000 mrem ** that are Tocated within large areas, such9 s PWR contair, ment, where no enclosure exists for purposes of locking, and no ediosure can be reasonably constructed around the individual areas, then that area [shall be roged off, conspicuously posted and a flashing light shall be activated as a warning device.

In lieu of the stay time specification of t5 (such as use of closed circuit TV camet As) continuous,e PEP, direct,or remote '

surveillance may be made by personnel qualified in radiation protection procedures to provide positive'.

exposure control over the a'ctivities'within the area.

5 6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 The PCP shall bejapproved.by sthe Commission prior to implementation.#

y s

6.13.2 Licensee initiated changes to the PCP:

(*

+

\\

1.

Shall be submitted to the Commission in the semi-annual Radioactive Effluent Release Report fd,the period in which the dadge(s) was t

made.

ThigsubmittalshalIcontain:

Sufficientlyf detailed information to totaNy support the rat'jonale a.

for the change without benefit of additional'.or supplemental 1

information;

%(

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t' o

b.

A determinayfon that the change did not reducp the overall l 2~

conformance of the solidified waste product,to existing criteria i

for solid wastes; and 4,

[(

it Documentationofthefactthat.thechangehasbeen\\reviewedand c.

found acceptable by the OSAC.'

1

/

q

- 2.

Shall become effective upon r'eview and acceptanc'e by the OSRC.

d v

- Measurement made at 18" from source of radioactivity.

The PCP shall be submitted and approved prior to shipment of ' wet" solid

, radioactive waste.

v e

SAN ONOFRE-UNIT 2 6-24 Amer.M ent Nn. 4

?

3 l

a.

,l'

7.0 SPECIAL TEST PROGRAM 7.1 For conducting the special low power test program as described in Sec-4 tion 22.2-1.G.1 of Supplement No. I to the Safety Evaluation Report (SER) the Technical Specifications may be exempt (E) or modified (C) as follows:

Technical Specifications Test Test Test I

Sect' ion Description Al A2 A3 2.2.I Reactor Trip Setpoints t

}.'

2.-

Linear Power Level-High C(1)

C(l)

C(1)

Four Reactor Coolant Pumps Operating 3.

Logarithmic Power Level-High C(2)

C(2)

C(2) 5.

Pressurizer Pressure-Low C(3) 7.

Steam Gen. Pressure-Low C(4)

C(4)

C(4) 9.

Local Power Density-High E(5)

E(5)

E(5) 10.

DNBR-Low E(5)

E(5)

E(5) 11.

Reactor Coolant Flow-Low E(5)

E(5)

E(5) 4 3.3.1 Reactor Protective Instrumentation l

9.

Local Power Density-High E(5)

E(5)

E(5) 1 10.

DNBR-Low E(5)

E(5)

E(5) 14.

Core Protection Calculators E(5)

E(5)

E(5) 16.

Reactor Coolant Flow-Low E(5)

E(5)

E(5) 3.3.2 Enginee~ red Safety Feature Actuation System Instrumentation j

~ '.

Safety Injection (SIAS)

C(3)

I l

i 4.

Main Steam Line Isolation C(4)

C(4)

L 6.

Containment Cooling (CCAS)

C(3)

!g 8.

Emergency Feedwater (EFAS)

C(4)

C(4)-

i i

i e

U s

l l

't 3'

\\

I

?

l 1

i SAN ONOFRE - UNIT 2 7-1 Amendment No. 4 l

l

Notes:

1.

Trip setpoint lowered to 5 9.1% RATED THERMAL POWER, allowable value i 10.4% RATED THERMAL POWER 2.

Trip setpoint raised to < 100% RATED THERMAL POWER, allowable value i 100% RATED THERMAL POWER 3.

Trip setpoint lowered to > 1,550 psia 4.

Trip setpoint lowered to 3 550 psia 5.

Trip bypassed

/

SAN ONOFRE - UNIT 2 7-2 Amendment No. 4

SAFETY EVALUATION AMENDMENT 4 to NPF-10 SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 2 DOCKET NO. 50-361 Introduction By letter dated May 14, 1982, the Southern California Edison Company (SCE), on behalf of itself, San Diego Gas and Electric Company, The City of Riverside and The City of Anaheim (the Licensees), requested the following changes to the San Onofre Nuclear Generating Station, Unit 2 Technical Specifications (TS).

1.

Proposed Change The Licensee proposes to modify Table 3.8-2, MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION BYPASS DEVICES. This proposed change will add three valves (HV-9377, 9378, and 9235) to the list, correct two valve numbers (LV-0227C and HV-8152), and modify the listed functional description of the valves to agree with the P&ID's and the instrument index.

Staff Evaluation This change will add three valves that were inadvertently omitted from the valve listing, correct some typographical errors, and more clearly identify the function of the valves. This change is approved because it corrects errors and improves the consistency of the Technical Specifications.

2.

Proposed Changes The Licensee proposed the following typographical changes:

a.

Technical Specification 3.4.8.3.1, APPLICABILITY:

change from "one any" to "any one",

b.

Technical Specification 3.7.1.5, Modes 2 and 3:

The word "in" was omitted.

It should read "Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ...."

c.

Technical Specification 3.10.5, Table 3.10-1, item 2.a: Main Steam Line Area Monitor "2RT-784781" should be "2RT-787481".

d.

Technical Specification 3.10.6, Li'ne 1:

the word "of" is misspelled as "fo" and should be corrected.

Staff Evaluation These changes are typographical in nature and therefore are approved.

San Onofre OL Amdt 4 1-1

3.

Proposed Changes The licensee has proposed the following modifications to the BASES Sections:

a.

Section 3/4.7.8 Fire Supression Systems:

insert a new paragraph.

"The San Onofre Unit 2 & 3 fire pumps and water supplies supply water to the San Onofre Unit I fire system.

Satisfactory completion of the Unit 2

& 3 fire pump and water supply surveillance requirements automatically satisfies the Unit 1 fire water supply requirements."

b.

Section 3/4.8, Add the following sentence to the discussion on fuel oil sampling.

" Reg. Guide 1.137 recommends testing of fuel oil samples in accordance with ASTM-0270-1975.

However, ASTM-D270-1965 was reverified in 1975, rather than re-issued.

The reverified 1965 standard is therefore the appropriate standard to be used."

Staff Evaluation These changes are for clarification of the Bases Sections and have no safety significance and are therefore approved, except for the second sentence under a., which will not be included because it is inappropriate to discuss Unit I requirements in the Unit 2 Technical Specifications.

4.

Proposed Changes In Technical Specifications Sections 3.1.2.7.b.3 and 3.1.2.8.b.3, the licensee proposes to change the required refueling water stcrage tank temperature from i

"a solution temperature between 40F and 120F" to "a solution temperature be-tween 40F and 100F, " for consistency with Technical Specification 3/4.5.4.

Staff Evaluation This is a change in the direction of greater conservatism since it restricts the allowable temperature range. The change also results in greater consist-ency between sections and is therefore approved.

5.

Proposed Cha'nges The licensee proposes to change Technical Specification 3.1.3.7 to allow the Part Length CEA group to be withdrawn to greater than or equal to 145", for consistency with Specification 3.1.3.4 and 3.1.3.5.

Staff Evaluation This change allows the Part Length CEA's to be considered " full out" at greater than or equal to 145", the same as the Shutdown CEA's of Specification 3.1.3.5.

This improves consistency between specifications and is therefore approved.

San Onofre OL Amdt 4 1-2

6.

Proposed Change The licensee proposes to change Technical Specification 3.3.1, Table 3.3-1, notation (c) to read " bypass shall be automatically removed when THERMAL POWER is greater than or equal to 5% of RATED THERMAL POWER", vice 1%, to be con-sistent with the operation of the bistable.

Table 3.3-1 ACTION 2 Line 7 reads,

" Specification 6.5.1.6k" and should read "6.5.1.6e" because parts of Section 6.5.1.6 were deleted and iten "k" was relettered as item "e".

Staff Evaluation The changes are typographical in nature and are approved.

7.

Proposed Change The Licensee proposes to change Technical Specification 3.3.2, Table 3.3-3, item 2, to delete the containment spray OPERABILITY requirement for Mode 4.

Containment Spray (CSAS) is required to be OPERABLE in Modes 1, 2, and 3 but not Mode 4 as defined in Technical Specification 3.6.2.1.

Staff Evaluation The containment spray system uses some of the components of the shutdown cooling system. Many of these components can be aligned only to one system at a time, either the containment spray system or the SDCS.

Since the shutdown cooling system is required to be operable in Mode 4, the containment spray system can not be operable also.

Therefore, this change is approved.

8.

Proposed Change The Licensee proposes to change Technical Specification 3.3.2, Table 3.3-3 Item 5 to be consistent with the implied operability requirement of Technical Specification 3.5.3 which recognizes that in addition, MODE 4 is applicable.

Therefore, Mode 4 should be added to Item 5 Table 3.3-3 under Applicable Modes.

Staff Evaluation 1

This additional operability requirement improves consistency between the specifications and is approved.

9.

Proposed Change The licensee proposes to change Technical Specification 3.3.2, Table 3.3-3 notation "a" to state " bypass shall be automatically removed when pressurizer pressure is greater than or equal to 400 psia", vice 500 psia.

This change will make Table 3.3-3 consistent with Technical Specification 3.3.1, Table 3.3-1 notation "b".

Staff Evaluation l

This change is typographical in nature and is approved.

l l

l San Onofre OL Amdt 4 1-3

10.

Proposed Change The Licensee proposes to change Technical Specification 3.3.2, Table 3.3-4, to delete Item 5a Manual (RAS). There are no manual RAS (Trip Buttons) in the plant and therefore this should be deleted.

Item 5 should be relettered as applicable.

Staff Evaluation There are no manual RAS (Trip Buttons) nor is there a requirement to install them. Therefore, this change is approved.

11.

Proposed Change The Licensee proposes to delete the following notation from Specification 3.4.1.3 "With the Reactor Coolant System cold leg temperature less than or equal to 235F, the SDCS isolation valves HV-9337, HV-9339, HV-9377, and HV-9378 shall be open with the SDCS relief valve PSV-9349 OPERABLE."

Staff Evaluation Specification 3.4.1.3 is the controling specification for the Reactor Coolant System in Mode 4.

The Specification for the Shutdown Cooling System is 3/4.4.8.3.1 and it specifies when the SDCS isolation valves should be open.

This statement in Specification 3.4.1.3 is redundant and the change is there-fore approved.

12.

Proposed Change The Licensee proposes to delete the words "with all suction line valves open" from Specification 3.4.1.4.1.

Staff Evaluation This request is similar to the above request in item 11 in that the Shutdown Cooling System suction valves are controlled by the SDCS Specification.

Therefore this change is approved.

13.

Proposed Change The Licensee proposes to add the words " powered from the IE busses" to Specifi-cation 4.4.3.2.

Staff Evaluation This change is intended to clarify the pressurizer heater power supply require-ment and is in accordance with SER Supplement No. 1, Section 22.2-II.E.3.1.

We concur with the proposed change but will add the required words to Specifica-tion 3.4.3.

San Onofre OL Amdt 4 1-4

covering the company organization or contractor organization cognizant of the work conducted under the procedure.

Staff Evaluation The proposed change to Technical Specification 6.5.2.2 adds a.new specification.

It is unacceptable because it would remove the responsibility for procedure preparation and review and review of nuclear safetyrelated programs, required elsewhere in the technical specifications, from the plant management to the quality assurance program.

QA programs are not set up to address technical adequacy, but may be used for audits.

Without the new proposed Technical Specification 6.5.2.2, there is no need to renumber the succeeding specifications 6.5.2.3 through 6.5.2.10.

18.

Proposed Change Technical Specification 6.5.3.4.a:

The NSG shall review:

The safety evalua-tions for (1) changes to procedures required by Specification 6.8, equipment or systems and (2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.

Staff Evaluation The proposed change to Technical Specification 6.5.3.4.a merely clarifies the fact that the procedures referred to are only those required by Specification 6.8.

There is no safety significance to the proposed change and it is, there-fore, acceptable.

19.

Proposed Change Technical Specification 6.8.2:

Each procedure of 6.8.1 and changes thereto shall be approved by the Station Manager or by (1) the Manager, Operations (2) the Manager, Technical (3) the Manager, Maintenance (4) the Deputy Station Manager or (5) the Manager, Health Physics as previously designated by the Station Manager prior to implementation and shall be reviewed periodically as set forth in administrative procedures.

Staff Evaluation The proposed change to Technical Specification 6.8.2 merely removes the top-line management from having to review each procedure or change personally.

The top management still has the responsibility for the review, as required _by Specification 6.5.2.1, and still must approve the procedure or change.

The proposed change makes 6.8.2 consistent with 6.5.2.1, and is acceptable.

20.

Proposed _ Change The licensee proposes to delete the requirement of monthly sampling local vegetation and performing a gamma isotopic analysis on the sample.

San Onofre OL Amdt 4 1-6 i

14.

Proposed Change The following should be added to Specification 6.2.4:

"The Shift Technical Advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design and in the response and analysis of the plant for transients and accidents."

Staff Evaluation This addition is needed because, while the qualifications for all other plant staff are included in the Technical Specifications, those for the Shift Techni-cal Advisor were, inadvertently, not included.

15.

Proposed Change Change Technical Specification S 6.1-2, 6.2-2, Table 6.2-1, Figure 6.2-2 and Figure 6.2-3 to reflect current personnel titles.

Staff Evaluation

~

The proposed changes to Figure 6.2-2 are primarily changes in position titles, as is suggested by the Licensee's description of the change.

The inclusion of note 4 merely clarifies the fact that the Manager, Station Emergency Prepared-ness, is responsible for plant fire protection.

There is no safety signifi-cance in these changes or in the deletion of the administrative position of Supervisor, Planning and Budgeting, under the Deputy Station Manager.

There-fore, these changes are acceptable.

The proposed changes to Technical Specifications 6.1.2, 6.2.2.e, Table 6.2.-1 and Figure 6.2-3 are of no safety significance and are acceptable because they only change the " Watch Engineer" position title to " Shift Supervisor."

16.

Proposed Change Technical Specification 6.5.2.1:

Each procedure and program required by Specification 6.8 and changes thereto, shall be prepared by a qualified indi-r vidual/ organization.

Each such procedure, and changes thereto, shall be reviewed by an individual / group other than the individual / group which prepared the procedure, or change, thereto, but who may be from the same organization as the individual / group which prepared the procedure, or changes thereto.

l l

Staff Evaluation The proposed change to Technical Specification S 6.5.2.1 would delete the Station Manager's responsibility for assuring that the processes covered by 6.5.2.1 are carried out.

The Station Manager must retain this ultimate responsibility.

Therefore, the proposed change is unacceptable.

17.

Proposed Change Technical Specification 6.5.2.2:

Procedures and Programs which affect nuclear safety (but are not required by Specification 6.8) and changes thereto, shall be prepared and reviewed in accordance with the Quality Assurance Program San Onofre OL Amdt 4 1-5

4 Staff Evaluation The San Onofre area is classified as semi-arid and local vegetation does not grow most of the year.

Therefore, local vegetation is not necesarily a reliable indicator of an exposure pathway.

As an alternative, the licensee has added a statement to Section 5.0 of the Offsite Dose Calculation Manual which commits to sampling the milk of a milk producing animal found during land use census performed in accordance with Technical Specifications 3.12.2.

Sampling this milk will adequately cover this ingestion pathway.

For these reasons, this change is approved.

21.

Proposed Change Technical Specification 4.6.1.3a:

The Licensee proposed to change the door seal test pressure from 10 psig or greater to 9.5+0.5 psig.

This change will assure the proper seating of the door seal by verifying the seal leakage is less than or equal to 0.01La, which is consistent with industry standards, while remaining within the limits of the manufacturer's pressure recommenda-tions to avoid door seal damage.

Staff Evaluation Changing the door seal test pressure from 10 psig or greater to 9.5+0.5 psig is necessary to meet the door seal manufacturer's recommendation to prevent damage to the door seals.

The test pressure of 9.5 psig is as reliable as 10.0 psig for testing the proper seating of the air lock door seals.

Therefore, this change is acceptable and is approved.

22.

Proposed Change Add Section 7, Special Test Exceptions for Natural Circulation Tests to the Technical Specification for the duration of the special lower power test program as described in Section 22.2-I.G.1 of Supplement No. 1 to the Safety Evaluation Report, NUREG-0712.

Staff Evaluation In Supplement No. 1 to the SER the staff requirements and the licensees' commitments regarding special natural circulation testing were discussed.

The staff required that the licensee submit, four weeks prior to conducting the tests, detailed test procedures and a safety analysis.

By letter dated April 15, 1982, SCE provided the required information, thereby satisfying condition 2.B.(19)g of the San Onofre Unit 2 operating license, NPF-10, issued Februa ry 16, 1982.

The proposed natural circulation test was discussed with the licensees in a meeting, in Bethesda, Maryland, on May 20, 1982.

The staff has concluded that the proposed natural circulation tests are acceptable.

The basis for our conclusion is given in Supplement No. 6 to the San Onofre 2 and 3 Safety Evaluation Report, NUREG-0712, dated June 1982.

23.

Proposed Change Technical Specification 3.10.5, Table 3.10-1, item 5:

Change the second sentence to read " Containment airborne menitor 2RT-7804-1 or 2RT-7807-2 and San Onofre OL Amdt 4 1-7

associated sampling media shall perform these functions prior to initial criticality." This change corrects the Technical Specifications to agree with Supplement No. 5 to the SER, Section 11.3. _The Bases section corresponding to this Specification was also corrected.

Staff Evaluation This change corrects an error in the Technical Specifications. -The corrected wording agrees with the previously issued Safety Evaluation and is acceptable.

24.

Proposed Change In several separate discussions with-the' staff the licensee determined that the following Technical Specifications changes were necessary for the reasons noted.

1-8

~

Page Change Reason 3/4 1-20 1.

Change " Control" CEA to Standard Terminology

" Regulating CEA".

2.

Change action statement to Standard Terminology Hot Standby within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

3.

Change Surveillance require-Clarification ment to indicate that the requirement is in force after reinstallation of the reactor vessel head.

3/4 3-11 Add once per shift and Clarification not monthly surveillance on a new requirement.

CPC's 70% Pwr, 3/4 3-51 Change method of showing Clarification.

to 3-53A action requirements for inoperable accident monitoring instruments.

3/4 4-1 Change 3.4.1 to 3.4.1.1 Typo Change 4.4.1 to 4.4.1.1 Typo 3/4 4-31 Change Title of Section to PRESSURIZER-HEATUP/C00LDOWN Clarification 3/4 4-32 Add Relief Valve The overpressure and 4-33 Isolation Valve re-Protection System quirement to LCO.

Technical Specifica-l tion (T.S. 3.4.8.3) was split into two specifications; one applicable to RCS temperature 235 F, and one applicable to RCS 235 F.

This re-quirement was inadver-tently deleted in Technical Specifica-tion 3.4.8.3.2, RCS Temperature > 235 F and is being add for consistency of the re-quirement of the over-all specification.

San Onofre OL Amdt 4 1-9

i Page Change Reason 3/4 4-33 Add 30-day report.

and 4-33)

(Identical to 3/4 4-32 requirement to the action statement.

Change surveillance method.

3/4 6-9 Change Retensioning To make this require-Specification to be ment.in compliance in accordance with with the recommenda-Regulatory Guide of Proposed Revision 3 to Regulatory Guide 1.35,

" Determining Prestressing Force for Inspection of Prestressed Conrete Containments," April 1979, and Proposed Regulatory Guide 1.35.1,

" Inservice Surveil-lance of Ungrouted Tendons in< Pre-stressed Concrete Containment Structures," April 1979.-

3/4 6-10 Change terminology to agree with Regulatory Guide (Identical to 3/4 6-9) 3/4 6-11 Require complete (Identical to 3/4 6-9) grease coverage 3/4 6-12a Change method of noting Clarification Tendon Ends and Buttresses 3/4 6-23 Add # notation to Typo SDC Relief Valve i

3/4 7-4 Change exception of Clarification Specification 4.0.4 to show.that it only applies

'to the' turbine driven Auxiliary Feedwater Pump t

-3/4 10-3 Add " Table 2.2.-1" to This Special Test LC0 Exception cannot be performed without suspending the-requirement of l

' Table 2.2-1 San Onofre OL Amdt 4 -

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Environmental Consideration We have determined that this amendment does not authorize a change in effluent types or total amount nor an increase in power level and will not result in any significant environmental impact.

Having made this determination, we have further concluded that this amendment involves action which is insignificant from the standpoint of environmental impact, and, pursuant to 10 CFR Section 51.S(d)(4), that an environmental impact statement or negative declaration and environmental irpact appraisal neednot be prepared in connection with the issuance of this statement.

Conclusion Based upon our evaluation of the proposed changes to the San Onofre, Unit 2 Technical Specifications, we have concluded that:

(1) because this amendment does not involve a significant increase in the probability or consequences of accidents previously considered, does not create the possibility of an accident of a type different from any evaluated previously, and does not involve a significant decrease in a safety margin, this amendment does not involve a significant safety hazards consideration; (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

We, therefore, conclude that the proposed changes are acceptable.

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