ML20237G684

Amends 60 & 49 to Licenses NPF-10 & NPF-15,respectively, Revising Licenses by Deleting License Condition 2.C(5) Re Environ Qualification of Electrical Equipment & Revising Tech Specs Including 3/4.7.1.5, Msivs

ML20237G684
Person / Time
Site:	San Onofre
Issue date:	08/14/1987
From:	Knighton G Office of Nuclear Reactor Regulation
To:
Shared Package
ML20237G687	List: ML20237G684 ML20237G706
References
NUDOCS 8709020369
Download: ML20237G684 (28)

Text

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uq'o, UNITED STATES 8,

NUCLEAR REGULATORY COMMISSION o

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WASHINGTON, D. C. 20555

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SOUTHERN CALIFORNIA EDISON COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY i

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THE CITY OF RIVERSIDE, CALIFORNIA THE CITY OF ANAHEIM, CALIFORNIA DOCKET NO. 50-361 SAN _0NOFRE NUCLEAR GENERATING STATION, UNIT 2 AMENDMENT TO FACILITY OP_ERATING LICENSE Amendment No. 60 License No. NPF-10 1.

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

A.

The applications for amendment to the license for San Onofre Nuclear Generating Station, Unit 2 (the facility) 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 January 25, 1984, April 19, 1985, July 1, 1985, October 25, 1985, and February 7, 1986, as supplemented by letter dated September 6, 1986, comply with the standards and require-ments of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's regulations as set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the applications, 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 amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

8709020369 B70814 ADOCK0500g1 DR

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2 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this a'nendment and Paragraph 2.C(2) of Facility Operating License No. NPF-10 is hereby amended to read as follows:

(2) Technical Specifications l

The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 60, are tmreby incorporated in the license.

SCE shall operate the facility in accordarce with the Technical i

Specifications and the Environmental Protection Plan.

3.

In addition, the license is amended by changing Paragraph 2.C(5) of l

Facility Operating License NPF-10, which is hereby amended to read as l

follows:

l (5) Environmental Qualification.

This paragraph intentionally deleted.

i l

l 4

The chu.ge to Paragraph 2.C(5) cf the license is effective as of the date of issuance of this amendment. The changes in Technical Specifications are to become effective within 30 days of issuance of the amenknent.

In the period between issuer.ce of the amendment and the effective date of the new Technical Specifications, the licensees shall adhere to the Technical Specifications existing at the time. The period of time during changeover shall be minimized.

FOR THE NUCLEAR REGULATORY COMMISSION p % George W. Knighton, Director Project Directorate V Division of Reactor Projects - III, l

VI, V & Special Projects

Attachment:

Changes to the Technical Specifications Date of Issuance: August 14, 1937

.a

4 August 14, 1987

, I ATTACHMENT TO LICENSE AMENDMENT NO. 60 FACILITY OPERATING LICENSE NO. NPF-10 i

DOCKET NO. 50-361 i

I 1

Replace the fuilowing ] ages of the Appendix A (echnical Specifications with l

the enclosed pages. T1e revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Also to be replaced are the following overleaf pages'to the amended pages.

Amendrent Pages Overleaf Pages l

3/4 3-29 3/4 3-30 i

I 3/4 3-53 3/4 3-53a i

3/4 3-55 3/4 3-06 1

3/4 7-9 3/4 7-10

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I Table 3.3-5 (Continued)

INITIATING SIGNAL AND FUNCTION RESPONSE TIME (SEC)

)

5.

Steam Generator Pressure - Low l

MSIS (1) Main Steam Isolation (HV8204, HV8205)

~ ' 8. 9 l

(2) Main Feedwater Isolation (HV4048, HV4052) 10.9 (3) Steam, Blowdown and Sample Isolation 20.9 (HV8419,HV8421) j (HV4053, HV4054, HV4057, HV4058)

• (4) Auxiliary Feedwater Isolation (NOTE 7) 40.9

)

(HV4705,HV4713,HV4730,HV4731) 1 (HV4706,HV4712,HV4714,HV4715) 6.

Refueling Water Storage Tank - Low RAS (1) Containment Sump Valves Open,

50.7*

(2) ECCS Miniflow Isolation Valves Close 50.7* (Note 8) 7.

4.16 kv Emergcncy Bus Undervoltage l

LOV (loss of voltage and degraded voltage)

Figure 3.3-1 8.

Steam Generator Level - Low (and No Pressure-Low Trip)

EFAS (1) Auxiliary Feedwater (AC trains) 52.7*/52.7**

(2) Auxiliary Feedwater (Steam /DC train) 42.7 (NOTE 6) 9.

Steam Generator Level - low (and AP - High)

EFAS (1) Auxiliary Feedwater (AC trains) 52.7*/52.7**

(2) Auxiliary Feedwater (Steam /DC train) 42.7 (NOTE 6) 10.

Control Room Ventilation Airborne Radiation CRIS (1) Control Room Ventilation - Emergency Mode Not Applicable 11.

Control Room Toxic Gas (Chlorine)

TGIS (1) Control Room Ventilation - Isolation Mode 16 (NOTE 5)

12.,C.ontrol Room Toxic Gas (Ammonia)

TGIS Control Room Ventilation - Isolation Mode 36 (NOTE 5)

SAN ONOFRE-UNIT 2 3/4 3-29 AMENDMENT NO. 60 l

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Table 3.3-5 (Continued)

INITIATING SIGNAL AND FUNCTION RESPONSE TIME (SEC) 5.

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]

MSIS (1) Main Steam Isolation (HV8204, HV8205)

' ' 8. 9 l

(2) Main Feedwater Isolation (HV4048, HV4052) 10.9 (3) Stoam, Blowdown and Sample Isolation 20.9 (HV8419,HV8421)

(HV4053,HV4054,HV4057,HV4058)

• (4) Auxiliary Feedwater Isolation (NOTE 7) 40.9 (HV4705,HV4713,HV4730,HV4731)

(HV4706, HV4712, HV4714, HV4715) 6.

Refueling Water Storage Tank - Low RAS (1) Containment Sump Valves Open,

50.7*

(2) ECCS Miniflow Isolation Valves Close 50.7* (Note 8) 7.

4.16 kv Emergency Bus Undervoltage LOV (loss of voltage and degraded voltage)

Figure 3.3-1 8.

Steam Generator Level - Low (and No Pressure-Low Trip)

EFAS (1) Auxiliary Feedwater (AC trains) 52.7*/52.7**

(2) Auxiliary Feedwater (Steam /DC train) 42.7 (NOTE 6) j 9.

Steam Generator Level - Low (and AP - High)

EFAS (1) Auxiliary Feedwater (AC trains)

'52.7*/52.7**

(2) Auxiliary Feedwater (Steam /DC train) 42.7 (NOTE 6) l 10.

Control Room Ventilation Airborne Radiation CRIS (1) Control Room Ventilation - Emergency Mode Not Applicable 11.

Control Room Toxic Gas (Chlorine)

TGIS (1) Control Room Ventilation - Isolation Mode 16 (NOTE 5) 12.

Control Room Toxic Gas (Ammonia)

TGIS Control Room Ventilation - Isolation Mode 36 (NOTE 5)

I SAN ONDFRE - UNIT 2 3/4 3-29 AMENDMENT NO. 60 i

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Table 3.3-5 (Continued)

INITIATING SIGNAL AND FUNCTION RESPONSE TIME (SEC) 13.

Control Room Toxic Gas (Butane / Propane)

TGIS Control Room,entilation -

Isolation Mode 36 (NOTE 5) 14.

Fuel Handling Building Airborne Radiation FHIS Fuel Handling Building Post-Accident

• Cleanup Filter System Not Applicable

15. Containment Airborne Radiation CPIS Containment Purge Isolation 2 (NOTE 2) 16.

Containment Area Radiation CPIS Containment Purge Isolation 2 (NOTE 2)

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NOTES:

1.

Response times include movement of valves and attainment of pump or blower discharge pressure as applicable.

2.

Response time includes emergency diesel generator starting delay (applicable to A.C. motor-operated valves other than containment purge valves), instrumentation and logic response only.

Refer to Table 3.6-1 for containment isolation valve closure times.

3.

All CIAS actuated valves except MSIVs, MFIVs, and CCW E~aives 2HV-6211, 2HV-6216, 2HV-6223 and 2HV-6236.

4a. CCW noncritical loop isolation Valves 2HV-6212, 2HV-6213, 2HV-6218, and 2HV-6219 close.

4b.

Containment emergency cooler CCW isolation Valves 2HV-6366, 2HV-6367, 2HV-6568, 2HV-6369, 2HV-6370, 2HV-6371, 2HV-6372, and 2HV-6373 open.

5.

Response time inc'udes instrumentation, logic, and isolation damper closure times only.

6.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.

7.

Jnclude HV4762 and HV4763 following implementation of DCP 195J.

8 Prior to completion of DCP 6234, valve closure is manually initiated.

Following completion of DCP 6234, valves are to close automatically on a RAS coincident with a high-high containment sump signal.

Emergency diesel generator starting delay (10 sec.) and sequence loading delays'for SIAS are included.

Emergency diesel generator starting delay (10 sec.) is included.

SAN DNDFRE - UNIT 2 3/4 3-30 AMENDMENT NO. 55

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TABLE 3.3-10(Continuedj ACTION STATEMENTS ACTION 20 -

With the number of OPERAPLE accident monitor.ing 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 22 -

With the number of OPERABLE Channels orie less than the Required-Number of Channels, either restore the system to OPERABLE status within 7 days if repairs are feasible without shutting down or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within*30 days following the event out-lining the action.taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

ACTION 23 -

With the number of OPERABLE Channels one less than the Minimum Channels OPERABLE in Table 3.3-10, either restore the inoperable channel (s) to OPERABLE status within 48 hcurs if repairs are feasible without shutting down or:

1.

Initiate an alternate method of monitoring the reactor vessel inventory; 2.

Prepare and submit a Special Report to the Commission pur-suant to Specification 6.9.2 within 30 days following the event outlining the action taken, the cause of the inopera-bility and the plans and schedule for restoring the system-to OPERABLE status; and 3.

Restore both channels of the syster,to OPERABLE status at the next scheduled refueling.

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SAN ONOFRE-UNIT 2 3/4 3-53a AMENDMENT NO. 60

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INSTRUMENTATION FIRE DETECTION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.7 As a minimum, the fire detection instrumentation for eact; fire detection zone shown in Table 3.3-11 shall be OPERABLE.

APPLICABILITY: Whenever equipment protected by the fire detection instromant is required to be OPERABLE.

ACTION:

With the number of OPERABLE fire detection instrument (s) less than the minimum number OPERABLE requirement of Table 3.3-11:

a.

Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> establish a fire watch patrol to inspect the zone (s) l with the inoperable instrument (s) at least once per hqur, unless the l

instrument (s) is located inside the containment, then inspect the containment at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or monitor the containment air temperature at least once per hour at the locations listed in Specification 4.6.1.5.

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.7.1 Each of the above required fire detection instruments which are accessible during plant operation shall be demonstrated ORERABLE-at least once per 6 months by performance of a CHANNEL FUNCTIONAL TEST.

Fire detectors l

which are not accessible during plant operation shall be demonstrated OPERABLE by the performance of a CHANNEL FUNCTIONAL TEST during each COLD SHUTDOWN l

exceeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless performed in the previous 6 months.

I 4.3.3.7.2 The NFPA Standard 720 supervised circuits supervision associated with the detector alarms of each of the above required fire detection l

instruments shall be demonstrated OPERABLE at least once per 6 months.

l 4.3.3.7.3 The non-supervised circuits associated with detector alarms between l

the instruments and the control room shall be demonstrated OPERABLE at least once per 31 days.

4.'3. 3. 7. 4 Following a seismic event (basemat acceleration greater than or equal to 0.05 g):

a.

Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> each zone shown in Table 3.3-11 shall be inspected for fires, and f

b.

Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> an engineering evaluation shall be performed to verify the OPERABILITY of the fire detection system in each zone I

I shown in Table 3.3-11.

1 SAN ONOFRE-UNIT 2 3/4 3-56

PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION I

3.7.1.5 Each main steam line isolation' valve shall be OPERABLE.

APPLICABILITY:

MODES 1, 2 and 3.

ACTION:

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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; otherwise, reduce power to less than or equal to 5 percent RATED THERMAL POWER l

within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

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

l a.

The isolation valve is maintained closed.

I b.

The provisions of Specification 3.0.4 are not applicable.

l Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

l l

SURVEILLANCE REQUIREMENTS 1

4.7.1.5 Each main steam line isolation valve shall be demonstrated OPERABLE by verifying full closure within 8.0 seconds when tested parsuant to l

Specification 4.0.5.

SAN ONDFRE-UNIT 2 3/4 7-9 AMENDMENT NO 60

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l PLANT SYSTEMS 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION LIMITFHG 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 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.

e S

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

9

INSTRUMENTATION BASES i

room.

This capability is required in the event control room habitability is i

lost and is consistent with General Design Criteria 19 of 10 CFR 50.

The OPERABILITY of the remote shut'down instrumentation in Panel L411 ensures that sufficient capability is available to permit shutdown and mainte-nance of C,0LD SHUTDOWN of the facility in the event of a fire in the cable spreading room, control room or remote shutdown panel, L042.

3/4.3.3.6 ACCIDENT MONITORING INSTRUMENTATION 1

The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident.

This capability is consistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Plants to Assess Plant (*onditions During and Following an Accident," December 1975 and NUREG 0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations".

The containment high range area monitors (RV-148 & RU-149) and the main steamline radiation monitors (RU-139 A&B and RU-140 A&B) are in Table 3.3-6.

The high range effluent monitors and samplers (RU-142, RU-144 and RU-146) are in Table 3.3-13.

The containment hydrogen monitors are in Specification 3/4.6.5.1.

The Post Accident Sampling System (RCS coolant) is in Table 3.3-6.

The Subcooled Margin Monitor (SMM), the Heated Junctior. Thermocouple (HJTC),

and the Core Exit Thermocouple (CET) comprise the Inadequate Core Cooling (ICC) instrumentation required by Item II.F.2 NUREG-0737, the Post TMI-2 Action Plan.

The function of the ICC instrumentation is to enhance the ability of the plant operator to diagnose the approach to existence of, and recovery from ICC.,Addi-tionally, they aid in tracking reactor coolant inventory.

These instruments are included in the Technical Specifications at the request of NRC Generic Letter 83-37.

These are not required by the accident analysis, nor to bring the plant I

to Cold Shutdown.

In the event more than four sensors in a Reactor Vessel Level channel are inoperable, repairs may only be possible during the next refueling outage.

This is because the sensors are accessible only after the missile shield and reactor vessel, head are removed.

It is not feasible to repair a channel except during a refueling outage when the missile shield and reactor vessel head are removed to refuel the core.

If only one channel is inoperable, it should be restored to OPERABLE status in a refueling outage as soon as reasonably possible.

If both channels are inoperable, both channels shall be restored to OPERABLE status in the nearest refueling outage.

In the event that both HJTC channels are in-operable, existing plant instruments and operator training will be used as an alternate method of monitoring the reactor ves'sel inventory.

SAN ONOFRE - UNIT 2 B 3/4 3-3 AMENDMENT NO. 60

INSTRUMENTATION BASES 3/4.3.3.7 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires.

This capability is required in order to detect and locate fires in their early stages.

Prompt detection of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility fire protection program.

In the event that a portion of the fire detection instrumentation is inoperable, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the" inoperable instruments-tion is restored to OPERABILITY.

Since the fire detectors are non-seismic, a plant visual inspection for fires is required within two hours following.an earthquake (> 0.02g).

Since safe shutdown systems are protected by seismic Category I barriers rated at two and three hours, any fire after an earthquake should be detected by this inspection before safe shutdown systems would be affected.

Additionally, to verify the continued OPERABILITY of fire detection systems after an earthquake, an engineering evaluation of the fire detection instrumentation in the required zones is required to be performed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following an earthquake.

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3/4.3.3.8 RADI0 ACTIVE LIQUID EFFLUENT INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents.

The alarm /

trip setpoints for these instruments shall be calculated in accordance with the procedures in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.

The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.

3/4.3.3.9 RADI0 ACTIVE GASE0US EFFLUENT INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents.

The alarm / trip setpoints for these instruments shall be calculated in accordance with the procedures in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.

This instrumentation also includes provisions for monitoring and controlling the concentrations of potentially explosive gas mixtures in the waste gas holdup system.

The OPERABILITY and use of this instrumentation is consistent.with the requirements of General Design Criter.ia 60, 63 and 64 of Appendix,A to 10 CFR Part 50.

3/4.3.3.10 LOOSE-PART DETECTION INSTRUMENTATION The OPERABILITY of the loose part detection instrumentation ensures that sufficient capability is available to detect loose metallic parts in the pri-mary system and avoid or mitigate damar to primary system components.

The SAN ONOFRE - UNIT 2 B 3/4 3-4 AMENDMENT NO. 60 l

i

i INSTRUMENTATION BASES allowable out of-service times and surveillance requirements are consistent with the recommendations of Regulatory Guide 1.133, " Loose-Part Detection Pro-gram for the Primary System of Light-Water-Cooled Reactors," May 198L 3/4.3.4 TURBINE OVERSPEED PROTECTION This specification is provided to ensure that the turbine overspeed pro-l tection instrumentation and the turbine speed control valves are OPERABLE and i

will protect the turbine from excessive overspeed.

Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety l

related components, equipment or structures.

i i

t I

SAN ONOFRE - UNIT 2 B 3/4 3-5 AMENDMENT N0. 60

4

• * '!(pa ata ug'o UNITED STATES g

NUCLEAR REGULATORY COMMISSION o

[

i WASHINGTON, D C. 20555 y.....)

SOUTHERN CALIFORNIA EDISON COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY THE CITY OF RIVERSIDE, CALIFORNIA THE CITY OF ANAHEIM,' CALIFORNIA DOCKET NO. 50-362 SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 49 License No. NPF-15 1.

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

A.

The applications for amendment to the license for San Onofre Nuclear i

Generating Station, Unit 3 (the facility) 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 January 25, 1984, April 19, 1985, July 1, 1985, October 25, 1985, and February 7, 1986, as supplemented by letter dated September 6, 1986, comply with the standards and require-ments of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations as set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the applications, 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.

l l

, 2.

Accordingly, the. license is amended by changes to the Technical Specifications as indicated in the attachment to this amendment and Paragraph 2.C(2) of Facility Operating License No. NPF-15 is hereby amended to read as follows:

4 (2) Technical Specifications

. The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 49, are hereby incorporated in the license.

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

3.

In addition, the license is amended by changing Paragraph 2.C(5) of Facility Operating License NPF-15, which is hereby amended to read as follows:

(5) Environmental Qualification This paragraph intentionally deleted.

4.

The change to Paragraph 2.C(5) of the license is effective as of the date of issuance of this amendment. The changes in Technical Specifications are to become effective within 30 days of issuance of the amendment.

In the period between issuance of the amendment and the effective date of the new Technical Specifications, the licensees shall adhere to the Technical Specifications existing at the time. The period of time during changeover shall be minimized.

FOR THE NUCLEAR REGULATORY COMMISSION George W. Knighton, Director Project Directorate V Division of Reactor Projects - III, VI, V & Special Projects

Attachment:

Changes to the Technical Specifications Date of Issuance:

August 14, 1987

l August 14e 1987 i '

ATTACHMENT,TO L,ICENSE AMENDMENT NO. 49 FACILITY OPERATING LICENSE NO. NPF-15 DOCKET NO. 50-362 Replace 'the following pages of the Appendix A Technical Specificaticrs with the enclosed pages. The revised pages are identified by Amendnent number and contain vertical lines indicating the area of change. Also to be replaced are the following overleaf pages to the amended pages.

Amendment Page Overleaf Page 3/4 3-29 3/4 3-30 3/4 3-53 i

3/4 3-54 3/4 3-55 3/4 3-56 3/4 7-10 3/4 7-9 B 3/4 3-3 B 3/4 3-4 B 3/4 3-5

Table 3.3-5 (Continued) l INITIATING SIGNAL AND FUNCTION RESPONSE TIME (SEC)'

5.

Steam Generator Pressure - Low a.

MSIS (1) Main Steam Isolation (HV8204, HV8205) 8.9 (2) Main Feedwater Isolation (HV4048, HV4052) 10.9 l

(3) Steam, Blowdown and Sample Isolation 20.9 (HV8419,HV8421)

(HV4053,HV4054,HV4057,HV4058)

• (4) Auxiliary Feedwater Isolation (NOTE 7) 40.9 (HV4705,HV4713,HV4730,HV4731)

(HV4706,HV4712,HV4714,HV4715) 6.

Refueling Water Storage Tank - Low a.

RAS (1) Containment Sump Valves Open 50.7*

{

(2) ECCS Miniflow Isolation Valves Close 50.7* (Note 8) 7.

4.16 kV Emergency Bus Undervoltage a.

LOV (loss of voltage and degraded voltage)

Figure 3.3-1 8.

Steam Generator Level - Low (and No Pressure-Low Trip) a.

EFAS (1) Auxiliary Feedwater (AC trains) 52.7*/52.7**

(2) Auxiliary Feedwater (Steam /DC train) 42.7 (Note 6) 9.

Steam Generator Level - Low (and P - High) l a.

EFAS

)

(1) Auxiliary Feedwater (AC trains) 52.7*/52.7**

j (2) Auxiliary Feedwater (Steam /DC train) 42.7 (Note 6) 10.

Control Room Ventilation Airborne Radiation a.

CRIS l

(1) Control Room Ventilation - Emergency Mode Not Applicable l

(

11.

Control Room Toxic Gas (Chlorine) a.

TGIS (1). Control Room Ventilation. Iso.lation Mode 16 (NOTE 5)_

4 12.

Control Room Toxic Gas (Ammonia) a.

TGIS (1) Control Room Ventilation - Isolation Mode 36 (NOTE 5)

SAN ON0FRE - UNIT 3 3/4 3-29 AMENDMENT NO. 49' I

Table 3.3-5 (Continued)

INITIATING SIGNAL. AND FUNCTION RESPONSE TIME (SEC) 13.

Control Room Toxic Gas (Butane / Propane)

TGIS Control Room Ventilation -

Isolation Mode 36 (NOTE 5) 14.

Fuel Handlina Building Airborne Radiation FHIS Fuel Handling Building Post-Accident Cleanup Filter System Not Applicable

15. ' Containment Airborne Radiation CPIS Containment Purge Isolation 2 (NOTE 2) 16.

Containment Area Radiation CPIS Containment Purge Isolation 2 (NOTE 2)

NOTES:

1.

Response times include movement of valves and attainment of pump or blower discharge pressure as applicable.

2.

Response time incit, des emergency diesel generator starting delay (applicable to AC motor operated valves other than conTain~ merit purge valves), instrumentation and logic response only.

Refer to Table 3.6-1 for containment isolation valve closure times.

3.

All CIAS-Actuated valves except HSIVs and MFIVs and CCW valves 3HV-6211, 3HV-6216, 3HV-6223 and 3HV-6236.

4a.

CCW non-critical loop isolation valves 3HV-6212, 3HV-6213, 3HV-6218 and 3HV-6219.

4b.

Containment emergency cooler CCW isolation valves 3HV-6366, 3HV-6367, 3HV-6368, 3HV-6369, 3HV-6370, 3HV-6371, 3HV-6372 and 3HV-6373 open.

5.

Response time includes instrumentation, logic, and isolation damper closure times only.

6.

The provisions of Specification 4.0.4 are not applicable for entry into

. MODE 3.

7.

Include HV4762 and HV4763 following implementation of DCP 195J.

8.

Prior to completion of DCP 6234, valve closure is manually initiated.

Following completion of DCP 6234, valves are to close automatically on a RAS coincident with a high-high containment sump signal.

Emergency diesel generator starting delay (10 seconds) and sequence loading delays for SIAS are included.

Emergency diesel generator starting delay (10 seconds) is included.

j SAN ONOFRE - INIT.3 3/4 3-30 AMENDMENT NO. 44

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TABLE 3.3-10 (Continued)

ACTION STATEMENT _S ACTION 20 -

With the number of OPERABLE accident monitoring channels less than the Required Number of Channels, either restore the t

inoperable channel to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, j

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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT SHUTDOWN within the,next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

j ACTION 22 -

With the number of OPERABLE Channels one less than the Required Number of Channels, either restore the system to OPERABLE status within 7 days if repairs are feasible without shutting down or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event out-lining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

ACTION 23 -

With the number of OPERABLE Channels one less than the Minimum Channels OPERABLE in Table 3.3-10, either restore the inoperable channel (s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> if repairs are i

feasible without shutting down or:

1.

Initiate an alternate method of monitoring the reactor vessel inventory; 1

2.

Prepare and submit a Special Report to the Commission pur-l suant to Specification 6.9.2 within 30 days following,the l

event outlining the action taken, the cause of the inopera-bility and the plans and schedule for restoring the system i

l to OPERABLE status; and 3.

Restore both channels of the system to OPERABLE status at the next scheduled refueling, i

i 9

SAN ONOFRE - UNIT 3 3/4 3-54 AMENDMENT NO. 49

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1 l

4 SAN ONOFRE-t/ NIT 3 3/4 3-56 AMENDMENT NO. 20

TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC A.CTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF NE&SUREMENT SAMPLE AND ANALYSIS AND ANALYSIS FREQUENCY l

1.

Gross Activity Determination At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> j

2.

Isotopic Analysis for DOSE a)

I per 31 days, whenever the EQUIVALENT I-131 Concentration gross activity determination indicates iodine concentrations greater than 10% of the allowable limit.

b) 1 per 6 months, whenever the i

gross activity determination indicates iodine concentrations below 10% of the allowable limit.

1

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--_ -~-~~~ ~__- _

PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve shall be OPERABLE.

APPLICABILITY:

MODES 1, 2 and 3.

ACTION:

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

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

a.

The isolation valve is maintained closed.

b.

The provisions of Specification 3.0.4 are not applicable.

Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.1.5 Each main steam line isolation valve shall be demonstrated OPERABLE by verifying full closure within 8.0 seconds when tested pursuant to Specification 4.0.5.

l SAN ONOFRE-UNIT 3 3/4 7-10 AMENDMENT NO. 49

INSTRUMENTATION BASES room.

This capability is required in the event control roon habitability'is l

l lost and is consistent with General Design Criteria 19 of 10 CFR 50.

The OPERABILITY of the remote shutdown instrumentation in Panel L411 ensures that sufficient capability is available to permit shutdown and mainte-nance of C,0LD SHUTOOWN of the facility in the event of a fire in the cable spreading room, control room or remote shutdown panel, LO42.

3/4.3.3.6 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident.

This capability is consistent

~

with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Plants to Assess Plant Obnditions During and Following an Accident," December 1975 and NUREG 0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations".

The containment high range area monitors (RU-148 & RU-149) and the main steamline radiation monitors (RU-139 A&B and RU-140 A&B) are in-Table 3.3-6.

The high range effluent monitors and samplers (RU-142, RU-144 and RU-146) are i

in Table 3.3-13.

The containment hydrogen monitors are in Specification j

3/4.6.5.1.

The Post Accident Sampling System (RCS coolant) is in Table 3.3-6.)

The Subcooled Margin Monitor (SMM), the Heated Junction Thermocouple (HJTC),

and the Core Exit Thermocouple (CET) comprise the Inadequate Core Cooling (ICC) instrumentation required by Item II.F.2 NUREG-0737, the Post TMI-2 Action Plan.

The function of the ICC instrumentation is to enhance the ability of the plant operator to diagnose the approach to existence of, and recovery from ICC., Addi-tionally, they aid in tracking reactor coolant inventory.

These instruments are j

included in the Technical Specifications at the request of NRC Generic Letter 83-37.

These are not required by the accident analysis, nor to bring the plant to Cold Shutdown.

In the event more than four sensors in a Reactor Vessel Level channel are inoperable, repairs may only be possible during the next refueling outage.

This is because the sensors are accessible only after the missile shield and reactor vessel, head are removed.

It is not feasible to repair a channel except during a refueling outage when the missile shield and reactor vessel head are removed to refuel the core.

If only one channel is inoperable, it should be restored to OPERABLE status in a refueling outage as soon as reasonably possible.

If both channels are inoperable, both channels shall be restored to OPERABLE status in the nearest refueling outage.

In the event that both HJTC channels are in-operable, existing plant instruments and operator training will be used as an alternate method of monitoring the roactor vessel inventory.

SAN ONOFRE - UNIT 3 B 3/4 3-3 AMENDMENT N0. 49

INSTRUMENTATION BASES 3/4.3.3.7 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires.

This capability is required in order to detect and locate fires in their early stages.

Prompt detection of fires will reduce the potential for damage to safety-reiated equipment and is an integral element in the overall facility fire protection program.

In the event that a portion of the fire detection instrumentation is inoperable, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instruments-tion is restored to OPERABILITY.

Since the fire detectors are non-seismis, a plant visual inspection for fires is required within two hours following an earthquake (>0.02g).

Since safe shutdown systems are protected by seismic Category I barriers rated at two and three hours, any fire after an earthquake should be detected by this inspection before safe shutdown systems would be affected.

Additiona.lly, to verify the continued OPERABILITY of fire detection systems after an earthquake, an engineering evaluation of the fire detection instrumentation in the required zones is required to be performed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following an earthquake.

3/4.3.3.8 RADI0 ACTIVE LIQUID EFFLUENT INSTRUMENTATION i

The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents.

The alarm /

trip setpoints for these instruments shall be calculated in accordance with the procedures in the ODCH to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.

The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.

j i

3/4.3.3.9 RADI0 ACTIVE GASEOUS EFFLUENT INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents.

The alarm / trip setpoints for these instruments shall be calculated in accordance with the procedures in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.

This instrumentation also includes provisions for monitoring and controlling the concentrations of potentially explosive gas mixtures in the waste gas holdup system.

The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.

l I

l SAN ONOFRE - UNIT 3 B 3/4 3-4 AMENDMENT NO. 49

l 1

1 INSTRUMENTATION i

BASES 3/4.3.3.10 LOOSE-PART DETECTION INSTRUMENTATION The OPERABILITY of the loose part detection instrumentation ensures that sufficient capability is available to detect loose metallic parts in the pri-I mary system and avoid or mitigate damage to primary system components. The allowable out-of-service times and surveillance requirements are consistent with the r, recommendations of Regulatory Guide 1.133, " Loose-Part Detection Pro-gram for the Primary System of Light-Water-Cooled Reactors," May 1981.

3/4.3.4 TURBINE OVERSPEED PROTECTION

{

This specification is provided to ensure that the turbine overspeed pro-tection instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed.

Protection from tur-bine could generate potentially damaging missiles which could impact and damage l

safety related components, equipment or strudtures, i

4 I

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1 i

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SAN ON0FRE - UNIT 3 B 3/4 3-5 AMENDMENT NO. 49

__