Text

Corrected Amends 64 & 53 to Licenses NPF-10 & NPF-15, Respectively.Amends Change Tech Specs Re part-length Control Element Insertion Limits,Movable Control Assemblies,Safety Injection Tanks & ESFAS
	ML20205L048
Person / Time
Site:	San Onofre
Issue date:	08/03/1988
From:	Knighton G; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20205L051	List: ML20205L048;
References
	NUDOCS 8811010455
	Download: ML20205L048 (56)
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UNITED STATss e

3 NUCLE AR REGULATORY COMMISSION y

3 W A SHING TO N. D. C. eOSH 4 * *..

o' SOUTHERN CALIFORNIA EDIS0N COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY THE CITY OF RIVERSIDE, CALIFORNIA THE CITY OF ANAHEIM, CALIFORN!A DOCKET N0. 50 361, SAN ON0FRE NUCLEAR GENERATlhG STATION, UNIT 2 AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendeent No. 64 l

l License No, NPF 10 1.

The huclear Regulatory Comission (the Comission) has found that:

A.

The applications for amendment to the license for San Onofre Nuclea'r-Generating Station Unit 2 (the facility) filed by the Southern California Edison dorpany (SCE) on behalf of itself and San Diego Gas anc Electric Company, The City of Riverside and The City of Anaheim, California (licensees), dated April 27, 1984 and May 12 November 4, and Decerter 14, 1987 (as supplemented April 14 and May 6, 1988), comply with the standards and requirements of the Atomic Energy Act of 1954 as amended (the Act) and the Comission's regulationsassetforthIn10CFRChapterI; B.

The facility will operate in conformity with the applications, as CoeseissIon;e provisions of the Act, and the regulations of the amended th C.

There is reasonable assurance:

(1)thattheactivitiesauthorized by this amendment can be conducted without endangt ing the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Cosmission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this license amendment will not be inimical to the cossnon 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 Corsnission's regulations and all applicable requirements have been satisfied, fDu311010455 seico4 ADOCK 05000361 p

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2.

Accordingly, The license is amended by changes to the Technical Specifications as indicated in the attachment to this arendment and Paragraph J.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 Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 64, are hereby incorporated in the license.

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

3.

The changes in Technical Specifications are to become effective within 30 days of issuance of the amendment, with the exception of the change to Technical Specification 3/4.4.10, which is to become effective prior to startup from the Cycle 5 Refueling Outage.

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

4 This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGUL TORY Com!SSION

'fieorge hto[ Pro Director Project Directorate V Division of Reactor Projects. !!!,

1 IV, Y and Special Projects

Attachment:

Changes to the Technical Specifications Date of Issuance: August 3, 1988 4

a i

t e-ATTACHMENT 'O..!?_'.dE AMEN 0f4ENT NC 64

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FACILITY OPERATING LICENSE NO. NPF-1C DOCKET NO. 50-361 i

i Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by Amendrent number and contain vertical lines indicating the area of change. Also to be replaced are the following overleaf pages to the amended pages.

Arendment Pages Overleaf Pages l

Y xi xii 2-4 23

)

3/4 1 22 3/4 1 21 l

3/4 1 23 3/4 1-24 3/4 1-25 3/4 1-26 3/4 3-3 3/4 3-4

(

3/4 3-6 3/4 3-5 i

3/4 3-28 3/4 3-27 l

1 3/4 4 35 t

i 3/4 4-36 l

3/4 5-1 3/4 5-2 1

3/4 10-2 3/4 10-1 i

83/4 1 5 83/a 4-9 r

t l

l l

i t

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INDEX LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE HOTSRiJTD0W............................................

3/4 4 3 CO LD SHUTDOW - Loop s F t 11e d............................

3/4 4 5 COLD SHUTDOW - Loop s Not Fi l le d........................

3/4 4 6 3/4.4.2 S AF ETY VA LVE S - 0 P E RAT I NG...............................

3/4 4-7 3/4.4.3 PRES $uRIzER.............................................

3/4 4 8 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 OPERATIONAL LEAKAGE..................................

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

3/4 4 20 3/4.4.7 SPECIFIC ACTIVITY.......................................

3/4423.-

3/4.4.8 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM...............................

3/4427 PR E S SU R I Z E R - N E ATU P/C00 LDOW........................3/4 4 31 DVERPRESSURE PROTECTION SYSTEMS RCS TEMPERATURE < 235'F...........................

3/4 4 32 RCS TEMPERATURE I235'F.............................

3/4 4 33 3/4.4.9 ST R U C TU RA L I NT E G R I TY....................................3/4 4 34 3/4.4.10 REACTOR COOLANT GAS VENT SYSTEM.........................

3/4 4 35 l

3/4.5 EMERGENCY CORE C0OLING SYSTEMS 3/4.5.1 5AFETY IK1ECTION TANKS...............

3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T,,,1 350*F.......................... 3/4 5 3 3/4.5.3 ECCS SUBSYSTEMS - T,,,< 350*F..........................

3/4 5-7 3/4.3.4 REFUELING WATER STORAGE TANK............................

3/4 5 8 SAN ONOFRE UNIT 2 Y

AMENDMENT NO. 64

G o

INDEX BASES SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION.........

B 3/4 4 1 3/4.4.2 sArETY vAtvE5.........................................

3/4 4 1 3/4.4.3 n E 5 5U R u f R...........................................3/4 4 2 3/4.4.4 STEAM GENERATOR5........................'..............

B 3/4 4 2 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE........................

B 3/4 4 4 3/4.4.6 CHEMISTRY.............................................

B 3/4 4-4 3/4.4.7 SPECIFIC ACTIVITY.....................................

8 3/4 4 5 3/4.4.8 PRESSURE / TEMPERATURE LIMIT5...........................

8 3/4 4 6 3/4.4.9 STRUCTURAL INTEGRITY..................................

B 3/4 4 9 3/4.4.10 REACTOR COOLANT GAS VENT SY5 TEM.......................

B 3/4 4 9 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 SAFETY IKlECTION TANK5................................

B 3/4 5-1 3/4.5.2 and 3/4.5.3 EtCS SUBSYSTEMS...........................

B 3/4 5-1 3/4.5.4 REFUELI M WATER TANK..................................

B 3/4 5 2 t

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i SAN ONOFRE-UNIT 2 XI AMEN 0 MENT NO. 64

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t INDEX

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EASES SECTION PAGE 3/4.6 CONTAINENT SYST!5 3/4.6.1 PRIMARY CONTAI'. MENT......................

E 3/4 6-1 3/4.6.2 CEFRES$'JR:IAT :N A'.0 COCL:.3 SYSTEMS..................

E 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES...........'...............

2 3/4 6-4 3/4.6.4 CCMSUST! B LE G AS CONTR 0L...............................

B 3/4 6 5 l

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l SAN ONOFRE-UNIT 2 XII AMEN 0 MENT NO.16 l

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4 TABLE 2.2-1 h

REACTOR PROTECTIVE INSTRWENTATION TRIP SETPOINT LIMIT ~-

5,,

5 FUNCT10m4L tRIIT TRIP SETPOINT Att0WA8tE VALUES g

1.

Meneel Reacter Trip Not Applicable Not Appifcable

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2.

Lineer Peuer Level - Nigh -

Four Reacter Coelant Pumps i 110.0E of RATED THERMAL POWER i 111.3% of RATED THERML POWER Operating 3.

tegoritiumic Power Level - Nigh (1) 10.89E of RATED THERmt POWER i 0.96% of RATED THERMAL POWER 4.

Pressurizer Pressure - Nigh 1 2382 psia 1 2389 psia 5.

Pressurizer Pressure - Low (2) 1 1806 psia 1 1763 psia 3

6.

Contaisement Pressure - Nigh 1 2.95 psig i 3.14 psfg 7.

Steau Cenerater Pressure - Low (3)

> 729 psia

> 711 psia 8.

Steam Generator Level - Low

> 25% (4)

>24.23% (4) 9.

Local Power Density - Nigh (5) 1 21.0 kw/ft i 21.0 kw/ft l

10. DN8R - Low

> 1.31 (5)

11. Reacter Coelant Flow - Low I

a) DN Rate

< 0.22 psid/sec (6)(8)

< 0.231 psid/sec (6)(8) 9 b) Floor 5 13.2 psid (6)(8)

I 12.1 psid (6)(8) y c) Step i6.82psid(6)(8) 57.231psid(6)(8) x

12. Steam Generator Level - High 5 90E (4) i 90.74% (4)

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13. Selselc - High 5 0.48/0.60 (7) 1 0.48/0.60 (7)

~

14. loss of toad Turbine stop 'velve closed Turbine stop valve closed

9 TABLE 2.2-1 (Continued) m E

REACTOR PROTECTIVE INTRUMENTATION TRIP SETPOINT LIMITS o5g TABLE NOTATION my (1) Trip may be manually bypassed ateve 10 4% of RATED THERMAL POWER; bypass shall be automatically removed c;

when THERMAL POWER is less than or equal to 10 4% of RATED THERMAL POWER.

(2) Value may be decreased manually, to a minimum value of 300 psia, as pressurizer pre-s reduced, provided the margin between the pressurizer pressure and this value is maintained a 1<

, than or equal to 400 psi; the setpoint sha*1 be increased automatically as presturizer pressure i s.xreased until the trip setpoint is reached.

Trip may be manually bypassed below 400 psia; bypass,nall be automatically removed whenever pressurizer pressure is greater than or equal to 500 psia.

(3) Value may be decreased manually as steam generator pressure is reduced, provided the margin between the steas generator pressure and this value is maintained at less than or equc1 to 200 psi; the setpoint shall be increased automatically as steam generator pressure is increased until the trip setpoint is reached.

l-(4),. of the distance between steam generator upper and low level instrument nozzles.

(5) As stored within the Core Protection Calculator (CPC). Calculation of the trip setpoint includes measurement, calculational and processor uncertainties, and dynamic allowances. Trip may be manually bypassed below 10 4% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is greater than or equal to 10 4% of RATEC THERMAL POWER.

The approved DNBR 1!ait accounting for use of HID-2 grids is 1.31.

The byp y setpoint may be changed during ?? sting pursuant to Special Test l

Exception 3.10.2.

(6) DN RATE is the maximum decrease rate of the trip setpoint.

FLOOR is tse ainimum value of the trip setpoint, 5

STEP is the amount by which the trip -e'. point is below Me input signal unless limiter by DN kate j5 Wloor.

2 2

(7) Acceleration, horizontal / vertical, g.

8 (8) Latpo:.1t may be altered to disable trip functic.. during testing pursuant to Specification 3.10.3.

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REACTIVITY CONTROL SYSTEMS i

SHUTDOWN CEA INSERTION LIMIT LIMITING CONDITION FOR OPERATION Wh a 3.1.3.5 All. shutdown CEAs shall be withdrawn to greater than or equal to 145 inches.

APQTeABILITY: MODES 1 and 2*#.

ACTI..<.

f With a maximum of one shutdown CEA withdrawn to less than 145 inches, except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hour either:

Withdraw the CEA to greater than or equal to 145 inches, or a.

b.

Declare the CEA inoperable and apply Specification 3.1.3.1.

4 SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown CEA shall be determined to be withdrawn to greater than or equal to 145 inches:

Within 15 hinutes prior to withdrawel of any CEAs in regulating a.

groups during an approach to reactor criticality, and b.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereaf ter.

"See Special Test Exception 3.10.2.

With K,ff greater than or equal to 1.0.

SAN ONDFRE-UNIT 2 3/4 1-21

o REACTIVITY CONTROL SYSTEMS i

REGULATING CEA INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1. 3. 6 a.

When COLSS is in-service, the regulating CEA groups shall be limited to the withdrawal sequence and to the insertion limits shown on Figure 3.1-2.

The CEA insertion between the Long Term Steady State Insertion Limits and the Transient Insertion Limits is restricted to:

1.

Less than or equal to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval, 2.

Less than or equal to 5 Effective Full Power Days per 30 Effective Full Power Day Interval, and 3.

Less than or equal to 14 Effective Full Power Days per 365 Effective Full Power Day Interval, b.

When COLSS is cut-of-service, the regulating CEA groups shall be limited to the Short Term Steady State Insertion Limit shown on Figure 3.1-2.

The CEA insertion between the Long Term Steady State Insertion Limits and the Short Tern Steady State Insertion Limits is restricted to:

1.

Less than or equal to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval, 2.

Less than or equal to 5 Effective Full Power Days per 30 Effective Full Power Day Interval. and 3.

Less than or cqual to 14 Effective Full Power Days per 365 Effective Full Power Day Interval.

APPLICABILITY:

MODES 1* and 2*#,

ACTION:

When COLSS is in service and a.

With the regulating CEA groups inserted beyond the Transient Insertion Limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, within two hours either:

1.

Restore the regulating CEA groups to within the limits, or 2.

Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the CEA group position using the above figure.

"See Special Test Exceptions 3.10.2 and 3.10.4.

With K,ff greater than or equal to 1.0.

SAN ON0FRE-UNIT 2 3/4 1-22 AMENDMENT NO. 64

REACTIVliY CONTROL SYSTEMS ACTION: (Continued) t With the regulating CEA groups inserted between the Long Term Steady b.

Statr' Insertion Limits and the Transient Insertion Limits for intervals greater than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval, operation may proceed provided either:

1.

The Short Term Steady State Insertion Liuits of Figure 3.1-2 are not exceeded, or 2.

Any subsequent increase in THERMAL POWER is restricted to less than or equal to 5% of RATED THERMAL NMER per hour.

With the regulating CEA groups inserted between the Long Ters Steady c.

State Insertion Limits and the Transient Insertioa Limits for intervals greater than 5 EFPD per 30 EFPD interval or greater than i

14 EFPD per 365 EFPD intervat, either:

1.

Restore the regulating groups to within the Long Ters Steady State Inhertion Limits within two hours, or 2.

Se in at least HOT STAhDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

When COLSS is out of service and the regulating CEA groups are inserted beyond the Short Tern Steady State Insertion Limit except for Surveillance testing pursuant to Specification 4.1.3.1.2, within two hours either:

t Restore the regulating CEA group to within the limit, or a.

b.

Reduce thersai power to less than or equal to the fraction ef Rated l

Thermal Power which is allowed by the CEA group position and the Short Tern Steady State Insertion Limit.

SURVEILLANCE REQUIREMENTS 1

4.1.3.6 The position of each regulating CEA group shall be determined to be within the Transient Insertion Limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during i

time intervals when the PDIL Auctioneer Alare Circuit is inoperable, then verify the individual CEA positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The accumulated times during which the regulating CEA groups are inserted beyond the Long Ters Steady State Insertion Limits but within the Transient Insertion Limits shall be determined at Isast once per 24 hcurs.

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s SAN ONOFRE UNIT 2 3/4 1-23 AMENDMENT NO. 64 i

REACTIVITY CONTROL SYSTEMS

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SAN ONOFi.E-UNIT 2 3/4 1-24 AMENDMENT NO. 64

REACTIVITY CONTROL SYSTEMS oART LENGTH CEA INSERTION LIMITS LIMITING CONDITIOR FOR 0.P_ERATION E

The pa[t' length CEA group shall be limited to the insertion limits 3.1.3.7 shown on Figure 3.1-3 with PLCEA insertion between the Long Term Steady State Insertion Limit and the Transient Insertion Limit restricted to:

< 7 Effective Full Fower Days per 30 Effective Full Power Day a.

Tnterval,and

< i4 Effective Full Power Days per 365 Effective Full Pcwer Day b.

Tnterval.

APPLICABILITY: MODE I above 20% of RATED THERMAL POWER

ACTION:

a.

With the part length CEA groups inserted beyond the Transient Insera tion Limit, except for surveillance testing pursuant to specification 4.1.3.1.2, within two hours either:

1.

Restore the part length CEA groups to within the limit, or 2.

Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the PLCEA group position using Figure 3.1-3.

b.

With the part length CEA groups inserted between the Long Term Steady State Insertion Limit and the Transient Insertion Limit for intervals

> 7 EFPD per 30 EFPD interval or > 14 EFPD per 365 EFPD interval, either:

1.

Restore the part length groups to within the Long Tern Steady State Insertion Limit within two hours, or 2.

Reduce THERMAL POWER to less than or equal to 20% of RATED THERMAL IWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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

The accumu-lateo time during which the part length CEA groups are inserted beyond the Long Tern Steady State Insertion Limit but within the Transient Insertion Limit shall be deterwirad at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

See Special Test Exception 3.10.2.

SAN ONOFRE-UNIT 2 3/4 1-25 AMENDMENT NO. 64

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TABLE 3.3-1 y

REACTOR PROTECTIVE INSTRUMENTATION o

MINIMUM 5

TOTAL NO.

CHANNELS CHANNELS APPLICABLE y

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION E-1.

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

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

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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

g.

Pressurizer Pressure - High 4

2 3

1, 2 2#, 3#

5.

Pressurizer Pressure - Low 4

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

6.

Containment Pressure - High 4

2 3

1, 2 2#, 3#

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7.

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

J, 8.

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

9.

Local Power Density - High 4

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

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10. DNBR - Low 4

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

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

i2. Reactor Protection System Logic 4

2 3

1 2 2#, 3#

A 3, 4*, 5*

7A

13. Reacter Trip Breakers 4

2(f) 4 1, 2 5

3*, 4*, 5*

7A g

14. Core Protection Calculators 4

2(c)(d)(e) 3 1, 2 2#, 35, 7 l

g

15. CEA Calculators 2

1 2(e) 1, 2 6#, 7 l

g

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

g

17. Seismic - High 4

2 3

1, 2 2#, 3#

18. Loss of Load 4

2 3

1(g) 2#, 3#

TABLE 3.3-1 (Continued)

TABLE NOTATION A

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 POW 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.

(c) Trip may be manually bypassed below 10'4% of RATED THERMAL POWER; bypass

Mll be aulg% of RATED THERMAL POWER.matically removed when THERMAL PO equal to 10 During testing pursuant to Special Test Exception 3.10.2 or 3.10.3, trip may be manually bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically rcmoved when THERMAL POWER is greater than or equal to 5% of AATED THERMAL POWER.

(d) Trip may be bypassed during testing purauant to Special Test Exception 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.

(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 channel to OPERABLE status withip 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 To'al Number of Channols, STARTUP and/or POWER OPERATION may cor.tinue 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 ;ater than during the next COLD SHUT 00W9 SAN ONOFRE-UNIT 2 3/4 3-4 AMENDMENT NO. 64

TABLE 3.3-1 (Centinued)

ACTION STATEMENTS With a channel process measurement circuit that affects multiple functional units inoperable or in test, bypass or trip all associated functional units as listed below:

Process Measurement Circuit Functional Unit Bypassed 1.

Linear Power Linear Power Level - High (Subchannel or Linear)

Local Power Density - High DNBR - Low 2.

Pressurizer Pressure - High Pressurizer Pressure - High Local Power Density - High DNBR - Low 3.

Containment Pressure - High Containment Pressure - High (RPS)

Containment Pressure - High (ESF) 4.

Steam Generator Pressure -

Steam Generator Pressure - Low Low Steam Generator AP 1 and 2 (EFAS 1 and 2) 5.

Steam Generator Level Steam Generator Level - Low Steam Generator Level - High Steam Generator AP (EFAS) 6.

Core Protection Calculator Local Power Density - High DNBR - Low ACTION 3 -

With the. number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement, STARTUP and/or POWER OPERATION i

may continue provided the following conditions are satisfied:

a.

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

Linear Power Linear Power Level - High (Subchannel or Linear)

Local Power Density - High DN8R - Low SAN ONOFRE-UNIT 2 3/4 3 5

TABLE 3.3-1 (Continued)

ACTION STATEMENTS 2.

Pressurizer Pressure -

Pressurizer Pressure - High High Local Power Density - High DNBR - Low 3.

Containment Pressure -

Containment Pressure - High (RPS)

High Containment Pressure - High (ESF) 4.

Steam Generator Steam Generator Pressure - Low Pressure - Low Steam Generator AP 1 and 2 (EFAS 1 and 2) 5.

Steam Generator Level Steam Generator Level - Low Steam Generator Level - High Steam Generator AP (EFAS) 6.

Core Protection Local Power Density - High Calculator DNBR - Low 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 2 are satisfied.

ACTION 4 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes.

1 1

ACTION 5 With the number of channels OPERABLE one less than required by I

the Minimum Channels OPERABLE requirement, he in at least HOT STANDCY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTIO) 6 a.

With 'one CEAC inoperable, operation may continue for up to 7 days provided that at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , each CEA is verified to be within 7 inches (indicated position) of all other CEA's in its group.

After 7 days, operation may continue provided that Action 6.b is met.*

If the exemption to Specification 3.0.4 is used, Action 6.b must be met.

b.

With both CEACs inoperable, operation may continue provided that:*

1.

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the DNBR margin required by Specifica' tion 3.2.4.b (COLSS in service) or Specification 3.2.4.d (COLSS out of service) is satisfie.d.

Note:

Requirements for CEA position indication given in Technical Specification 3.1.3.2.

SAN ONOFRE-UNIT 2 3/4 3-6 AMENDMENT NO. 64

TABLE 3.3-5

_ ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME (SEC) 1.

Manual

^'

a.

SIAS Safety Injection Not Applicable Control Room Isolation Not Applicable Containment Isolation (3)

Net Applicable Containment Emergency Cooling Not Applicable b.

CSAS Containment Spray Not Applicable c.

CIAS Contairment Isolation Not Applicable d.

MSIS Main Steam Isolation Not Applicable e.

RAS Containment Sump Recirculation Not Applicable f.

CCAS Containment Emergency Cooling Not Applicable g.

EFAS Auxiliary Feedwater Not Applicable h.

CRIS Control Room Isolation Not Applicable 1.

TGIS Toxic Gas Isolation Not Applicable J.

FNIS Fuel Handling Building Isolation Not Applicable l

k.

CPIS Containment Purge Isolation Not Applicable SAN ONOFRE - UNIT 2 3/4 3-27

Table 3.3-5 (continued)

INITIATING SIGNAL AND FUNCTION RESPONSE TIME (SEC) 2.

Pressurizer Pressure-Low a.

SIAS (1) Safety Injection (a) High Pressure Safety Injection 31.2*

(b) Low Pressure Safety Injection 41.2*

(c) Charging Pumps 31.2*

(2) Control Room Isolation Not Applicable (3) Containment Isolation (NOTE 3) 11.2* (NOTE 2)

(4) Containment Spray (Pumps) 25.6*

(5) Containment Emergency Cooling (a) CCW Pumps 31.2*

(b) CCW Valves (Note 4b) 23.2*

(c) Emergency Cooling Fans 21.2*

3.

Containment Pressure-High a.

SIAS (1) Safety Injection (a) High Pressure Safety Injection 41.0*

(b) Low Pressure Safety Injection 41.0*

(2) Control Room Isolation Not Applicable (3) Containment Spray (Pumps) 25.4*

(4) Containment Emergency Cooling (a) CCW Pumps 31.0*

(b) CCW Valves (Note 4b) 23.0*

(c) Emergency Cooling Fans 21.0*

b.

CIAS (1) Containment Isolation 10.9* (NOTE 2)

(2) Main Feedwater Isolation 10.9 and Backup Isolation Valves (HV 4048, HV 4052, HV 1105, HV 1106, HV 4047, HV 4051)

(3) CCW Valves (Note 4a) 20.9 (4) Hainsteam Isolation Valves (HV 8204, 8.9 HV 8205)

(5) Minipurge Isolation Valves 5.9 4.

Containment Pressure - High-High CSAS Containment Spray 23.0*

SAN ONOFRE - UNIT 2 3/4 3-28 AMENDMENT NO. 64

)

REACTOR COOLANT SYSTEM 3/4.4.10 REACTOR COOLANT GAS VENT SYSTEM l

LIMITING CONDITION FOR OPERATION 3.4.10 The Rektor Coolant Gas Vent System shall be OPERABLE with; At least one of valves 2HV0296A or 2HV0296B ccpable of being powered a.

from an emergency bus and providing a vent path from the reactor vessel head; and, b.

At least one of valves 2HV0297A or 2HV02978 capable of being powered from an emergency bus and providing a vent path from the pressurizer steam space; and, At least one of valves 2HV0298, capable of being powered from an c.

emergency bus and provicing a vent path to the containment atmosphere, or 2HV0299, capable of being powered from an emergency bus and providing a vent path to the quench tank; and d.

Valves 2HV0296A, 2HV02968, 2HV0297A, 2HV02978, 2HV0299 and 2HV0298 i

all closed.

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

a.

With any of valves 2HV0296A, 2HV0296B, 2HV0297A or 2HV02978 inoperable, operation may continue provided that:

i) power is removed from the inoperable valve (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ;

and, ii) valves 2HV0299 and 2HV0298 are maintained closed and power is removed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and, iii) the inoperable valve (s) is restored to OPERABLE status during the next COLD SHUTDOWN.

b.

With any of valves 2HY0299 or 2HV0298 inoperable, operation may continue provided that:

1) power is removed from the inoperable valva (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ;

and, ii) valves 2HV0296A, 2HV0296B, 2HV0297A and 2HV02978 are all maintained closed and power is removed within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months sUnd SAN ONOFRE - UNIT 2 3/4 4-35 AMENDMENT NO.64 t

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION iii) the inoperable valve (s) is restored to OPERABLE status during the next COLD SHUTDOWN.

c.

The provisions of 3.0.4 are not applicable for entry into MODES 3, 2 and 1.

SURVEILLANCE REQUIREMENTS 4.4.10 Each reactor coolant system vent path shall be demonstrated OPERABLE at least once per 18 months by:

1.

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

2.

Cycling each valve in the vent path through at least one complete cycle of full travel from the control room during COLD SHUT 00WN or REFUELING.

3.

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

l l

SAN ONOFRE - UNIT 2 3/4 4-36 AMENDMENT NO. 64

}

e 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 SAFETY INJECTION TANXS LIMITING CONDITTON FOR OPERATION 3.5.1 Each reactor coolant system safety injection tank shall be OPERABLE with:

The, isolation valve open and power to the valve removed, a.

b.

A contained borated water volume of between 1680 and 1807 cubic

feet, Between 1850 and 2800 ppa of boron, and c.

d.

A nitrogen cover pressure of between 615 and 655 psia.

APPLICABILITY:

MODES 1, 2 and 3.*

ACTION:

With one safety injection tank inoperable, except as a result of a

~

a.

closed isolation valve, restore the inoperable 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , b.

With one safety injection tank inoperable due to the isolation valve being closed,ITANDBY within one hour and be in HOT SHUTDOWN w at least HOT the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.1 Each safety injection tank shall be demonstrated OPERA 8LE:

a.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:

1.

Verifying that the contained borated water volume and nitrogen cover-pressure in the tanks is within the above limits, and 2.

Verifying that each safety injection tank isolation valve is open.

l l

With pressurizer pressure greater than or equal to 715 psia.

SAN ONOFRE-UNIT 2 3/4 5-1 AMENDMENT NO.64 1

i

(

1

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b.

At least once per 31 days and within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months af ter each solution volume increase of greater +:ian or equal to 1% of tank volume by verifying the bo on conce..itration of the safety injection tank solution.

c.

At least once per ?,1 days by verifying the fuses removed from each safety injection cank vent valve.

d.

At least once per 31 days when the RCS pressure is above 715 psia, by verifying that the isolation valve optrator breakers are padlocked in the open position.

At least once per 18 months by verifying that each safety injection e.

tank isolation valve opens automatically under each of the following conditions:

1.

Before an actual or simulated RCS pressure signal exceeds 715 psia, and 2.

Upon receipt of an SIAS test signal.

I SAN CN0FRS-UNIT 2 34 52

O e

3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUT 00WN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for seasurement of CEA worth and shutdown margin provided reactivity equivalent to at least the highest estimated CEA worth is available for trip insertion from OPERABLE CEA(s).

APPLICABILITY: MODES 2 and 3*

ACTION:

a.

With any full length CEA not fully inserted and with less than the above reactivity equivalent available for trip insertion, immedi-ately initiate and continue beration at greater than or equal to 40 gpm of a solution containing greater than or equal to 2350 ppm l

boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored, b.

With all full length CEAs fully inserted and the reactor suberitical-by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 40 gpm of a solution containing greater than or equal to 2350 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

SURVEILLANCF REQUIREMENTS 4.10.1.1 The position of each full length and part length CEA required either partially or fully withdrawn shall be determined at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

4.10.1.2 Each CEA not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 7 days prior to reducing the SHUTDOWN E RGIN to less than the limits of Specification 3.1.1.1.

"Operation in MODE 3 shall be limited to 6 consecutive hours.

SAN ONOFRE-UNIT 2 3/4 10-1 AMENDMENT NO. 61

SPECIAL TEST EXCEPTIONS 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER OISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The moderator temperature coefficient, group height, insertion, and power distribution limits of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.1.3.7, 3.2.2, 3.2.3, 3.2.7, footnote C of Table 3.3-1 and footnote 5 of Table 2.2-1, and the Minimum Channels OPERABLE requirement of Functional Unit 15 of Table 3.3-1 may be suspended during the performance of PHYSICS TESTS provided:

The THERMAL POWER is restricted to the test power plateau which a.

shall not exceed 85% of RATED THERMAL POWER, and b.

The limits of Specification 3.2.1 are maintained and determined as specified in Specification 4.10.2.2 below.

APPLICABILITY:

MODES 1 and 2.

ACTION:

With any of the limits of Specification 3.2.1 being exceeded while the requirements of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.1.3.7, 3.2.2, 3.2.3, 3.2.7, footnote C of Table 3.3-1 and footnote 5 of Table 2.2-1, and the Minimum Channels OPERABLE requirement of Functional Unit 15 of Table 3.3-1 are suspended, either:

Reduce THERMAL POWER sufficiently to satisfy the requirements of a.

Specification 3.2.1, or b.

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

SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined at least once per hour during PHYSICS TESTS in which the requirements of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.1.3.7, 3.2.2, 3.2.3, 3.2.7, footnote C of Table 3.3-1 and footnote 5 of Table 2.2-1, or the Minimum Channels OPERABLE requirement of Functional Unit 15 of Table 3.3-1 are suspended ard shall be verified to be within the test power plateau.

4.10.2.2 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 Specifications 4.2.1.2 and 3.3.3.2 during PHYSICS TESTS above 5% of RATED THERMAL POWER in which the requirements of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.1.3.7, 3.2.2, 3.2.3, 3.2.7, footnote C of Table 3.3-1 and footnote 5 of Table 2.2-1, or the Minimum Channels OPERABLE requirement of Functional Unit 15 of Table 3.3-1 are suspended.

SAN ONOFRE-UNIT 2 3/4 10-2 AMENDMENT NO 64

REACTIVITY CONTROL SYSTEMS BASES MOVABLE CONTROL ASSEMBLIES (Continued)

The establishment of LSSS and LCOs require that the expected long and short ters behavior of the radial peaking factors be determined. The long term behavior relates to the variation of the steady state radial peaking factors with core burnup and is affected by the amount of CEA insertion assumed, the portion of a burnup cycle over which such insertion is assumed and the expected power level variation throughout the cycle.

The short tem behavior relates to transient perturbations to the steady-state radial peaks due to radial xenon redistribution.

The magnitudes of such perturtiations depend upon the expected use of the CEAs during anticipated power reduct'ons and load maneuvaring.

Analyses are performed based on the expected mode of operation of the NSSS (base load maneuvering, etc.) and from these analyses CEA insertions are deter-mined and a consistent set of radial peaking factors defined.

The Long Term Steady State and Short Term Insertion Limits are determined based upon the assumed mode of operation used in the analyses and provide a means of preserv-ing the assumptions on CEA insertions used.

The limits specified serve to limit the behavior of the radial peaking factors within tne bounds determined from analysir,.

The actions specified serve to limit the extent of radial menon-redistribution effects to those acconnodated in the analyses.

The Long and Short Term Insertion Limits of Specification 3.1.3.6 are specified for the plant which has been designed for primarily base loaded nperation but which has the ability to accosmodate a limited amount of load saneuvering.

The Transient Insertion Limits of Specification 3.1.3.6 and the Shutdown CEA Insertion Limits of Specification 3.1.3.5 ensure that 1) the minimum SHUT-DOWN MARGlH is maintained, and 2) the potential effects of a CEA ejection acci-dent are limited to acceptable levels.

Long ters operation at the Transient Insertion Limits is not permitted since such operation could have effects on the core power distribution which could invalidate esstaptions used to deter-eine the behavior of the radial peaking factors.

l The Part Length CEA Insertion Limits of Specification 3.1.3.7 ensure that adverse power shapes and rapid local power changes which affect radial peaking factors and DNS considerations do not occur as a result of a part length CEA group covering the same axial segesnt of the fuel assemblies for an extended period of time during operation.

l The CEA fully witurawn position is defined to be greater than or equal to 145 inches. The extreme limits of CEA travel, fully withdrawn and fully inserted, say be described as the upper electrical limit and lower electrical l

Ilmit respectively.

l SAN ONOFRE UNIT 2 5 3/4 1-5 AMEHOMENT NO.64

a j

REACTOR COOLANT SYSTEM 8ASE5 PRESSURE / TEMPERATURE LIMITS (Continued)

The OPERABILITY of the Shutdown Cooling Systen relief valve or a RCS vent opening of greater than 5.6 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 235'F.

The Shutdown Cooling System relief valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 100'F above the RCS cold leg te=peratures or (2) inadvertant safety injection actuation with two HPSI pumps injecting into a water-solid RCS with full charging capacity and letdown isolated.

3/4.4.9 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant.

These programs are in accordance with Section XI of the ASME Boiler and Presture Vessel Code and applicable Addenda,

i as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Coenission pursuant to 10 CFR Part 50.55a (g) (6) (1).

Components of the reactor coolant system were designed to provide access to permit inservice inspections in accordance with Section XI of the ASHE Boiler and Pressure Vessel Code,1974 Edition and Addenda through Summer 1975.

3/4.4.10 REACTOR COOLANT GAS VENT SYSTEM i

Reactor coolant systes gas vents are provided to exhaust noncondensible gases from the primary system that could inhibit aatural circulation core cooling following a non design bases accident.

The OPERASILITY of at least one reactor coolant systen vent path from the reactor vessel head and the pressurizer staan space ensures the capability exists to perform this function.

The design redundancy of the Reactor Coolant Gas Vent Systen serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, or control system does not prevent isolation of the vent path.

The function, capabilities, and testing requirements of the Reactor Coolant Gas Vent Systes are consistent with the requirements of Ites II,b.1 of MUREG-0737, "Clarification of TMI Action Plan Requirements," November 1980.

SAN ONOFRE-UNIT 2 5 3/4 4-9 AMENDMENT No,64

[

'o, UNITED STATES F

e NUCLEAR REGULATORY COMMISSION g

wAssiNoToN. o. c. rosss k..... p#

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 AMEN 0 MENT TO FACILITY ODERATING LICENSE Amendment No. 53 License No. NPF-15 1.

The Nuclear Regulatory Comission (the Comhsion) has fcund that:

A.

The applications for amendment to the license for San Onofre Nuclear Generating Station, Unit 3 (the facility) filed by the Southern California Edison Company (SCE) on behalf of itself and San Diego Gas and Electric Coepany, The City of Riverside 6nd The City of Anaheim, California (licensees), dated April 27, 1984 and May 12, Noventer 4, and December 14,1987 (as suppleinented April 14 and May 6, 1988), comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's regulatiors as set forth in 10 CFR Chapter I; B.

The facility will operate in confonnity with the applications, as amended, the provisions of the Act, and the regulations of the Comission; C.

There is reasonable assurance:

(1)thattheactivitiesauthorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in cospliance with the Coemission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this license amendment will not be inimical to the cosanon 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.

-2 2.

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

amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 53, are hereby incorporated in the license.

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

3.

The changes in Technical Specifications are to become effective within 30 days of issuance of the amendment.

In the period between isruance 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.

4 This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION torge W Knighton,[jectDirector Project irectorate Y Division of Reactor Projects - III, IV, V and Special Projects

Attachment:

Changes to the Technical Specifications Date of Issuance: August 3, 1988

a 5 !

ATTACHMENT TO LICENSE AMEtiDMENT NO. 53

~ ' '

FACILITY OPERATING LICENSE NO. NPF-15 DOCKET NO. 50-362 r

Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The 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.

Amendment page Overleaf page t

y vi l

xi xii 2-4 2-3 3/4 1-22 3/4 1-21 3/4 1-23 3/4 1-24 c

3/4 1-25 3/4 1-26 3/4 3-3 3/4 3-4 3/4 3-6 3/4 3-5 3/4 3-28 3/4 3-27 3/4 4-37 3/4 4-38 3/4 5-1 3/4 5-2 l

3/4 10-2 3/4 10-1 L

83/4 1-5 B3/4 4-10 B3/4 4-9 e

P L

I l

r t

i l

l t

INDEX i

LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION

_PAGE HOT SH0TDOWN............................................

3/4 4-3 COLD SHUTDOWN - LOOPS FI LLED............................

3/4 4-5 COLD SHUTD0YN - LOOPS NOT FI LLE0........................

3/4 4-6 3/4.4.2 SAF ETY VA LVE S - 0 PE RAT I NG...............................

3/4 4-7 3/4.4.3 PRESSURIZER.............................................

3/4 4-8 3/4.4.4 STEAM GENERATORS........................................

3/4 4-9 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE i

LEAKAGE DETECTION SYSTEMS............................

3/4 4-17 O P E RAT I O N A L L E A KAG E..................................

3/4 4-18 3/4.4.6 CHEMISTRY...............................................

3/4 4-21 3/4.4.7 SPECIF2C ACTIVITY.......................................

3/4 4-24

~

3/4.4.8 PRESSURE / TEMPERATURE LIMITS R E ACTO R COO LANT SYST EM............................... 3/4 4-28 P R E S SURIZ E R - MEATU P/C00 LDOWN........................

3/4 4-32 DVERPRESSURE PROTECTION SYSTEMS RCS TEMPERATURE < 285'F............................

3/4 4-33 RCS TEMPERATURE I 285'F............................

3/4 4 35 3/4.4.9 ST R U C IU RA L I NT EG R I TY....................................

3/4 4-36 3/4.4.10 REACTOR COO LANT GAS VENT SYSTEM.........................

3/4 4-37 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 1AF ETY I NJ E CT I ON T AN KS..................................

3/4 5-1 3/4.5.2 ECCS SUSSYSTEMS - T,,,1 350*F..........................

3/4 5 3 3/4.5.3 ECCS SUBSYSTEMS - T,,,< 350'F..........................

3/4 5-7 3/4.5.4 REFUELING WATER STORAGE TANK............................

3/4 5 8 SAN ONOFRE - UNIT 3 Y

AMENDMENT NO.53

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

3/4 6-1 CONTAI NMENT L E A KAG E..................................

3/4 6-2 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 VENTILATION SYSTEM.......................

3/4 6-14 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM.............................

3/4 6-15 IODINE REMOVAL SYSTEM................................

3/4 6-17 CONTAINMENT COOLING SYSTEM..........................

3/4 6-18 3/4.6.3 CONTAINMENT ISOLATION VALVES............................

3/4 6-19 3/4.6.4 COMBUSTIBLE GAS CONTROL HYDROGEN M0NITORS....................................

3/4 6-27 i

ELECTRIC NYOROGEN REC 0MBINERS........................

3/4 6-28 CONTAINMENT DOME AIR CIRCULAT0RS.....................

3/4 6-29 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES........................................

3/4 7-1 AUXILIARY FEEDWATER SYSTEM...........................

3/4 7-4 CONDENSATE STORAGE TANKS.............................

3/4 7-6 ACTIVITY.............................................

3/4 7-8 MAIN STEAM LINE ISOLATION VALVES.....................

3/4 7-10 i

SAN ONOFRE-UNIT 3 VI l

l 1

INDEX 8ASES SECTION pg 3/4.4.4 STEAM GENERATORS......................................

B 3/4 4-2 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE........................

B 3/4 4-4 3/4.4.6 CHEMISTRY.............................................

8 3/4 4-4 3/4.4.7 SPECIFIC ACTIVITY.....................................

B 3/4 4-5 3/4.4.8 PRESSURE / TEMPERATURE LIMITS...........................

8 3/4 4-6 3/4.4.9 STRUCTURAL INTEGRITY..................................

B 3/4 4-10 3/4.4.10 REACTOR COOLANT GAS VENT 5YSTEM.......................

B 3/4 4-10 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 SAFETY INJECTION TANKS................................

8 3/4 5-1 I

3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS.................'..........

B 3/4 5 1 3/4.5.4 REFUELING WATER STORAGE TANK..........................

B 3/4 5 2 t

3/4.6 CONTAIN!NMENT SYSTEMS 3/4.6.1 P R I M R Y CON TA I NM E NT...................................B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS..................

8 3/4 6 3 3/4.6.3 CONTAINMENT ISOLATION VALVES..........................

8 3/4 6-4 6

3/4.6.4 COMBUSTIBLE GAS CONTR0L...............................

8 3/4 6 5 1

l

[

f 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE.........................................

B 3/4 7-1 l

3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.......

8 3/4 7-3 3/4.7.3 COMPONENT COOLING WATER SYSTEM........................

8 3/4 7-3 3/4.7.4 SALT WATER COOLING SYSTEM.............................

8 3/4 7-3 3/4.7.5 CONTROL ROOM EMERGENCY AIR CLEANUP SYSTEM.............

8 3/4 7-4 t

3/4.7.6 5NUBBERS..............................................

8 3/4 7-5 i

3/4.7.7 SEALED SOURCE CONTAMINATION...........................

B 3/4 7-6 3/4.7.8 FIRE SUPPRESSION SYSTEMS..............................

B 3/4 7-6 3/4.7.9 FI RE RATED AS S EM LI E S.................................

B 3/4 7-7 r

r SAN ONOFRE - UNIT 3 Xi AMENDMENT NO.53

,r--

~ --

Lo-1 l

INDEX BASES SECTION PAGE 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2 and 3/4.8.3 A.C. SOURCES, D.C. SOURCES and ONSITE POWER DISTRIBUTION SYSTEMS................

E 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVI'.E5.............,

B 3/4 8-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION...................................

B 3/4 9 1 3/4.9.2 INSTRUMENTATION.......................................

B 3/4 9-1 3/4.9.3 DECAY TIME............................................

B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS.....................

B 3/4 9-1 3/4.9.5 COMMUNICATIONS........................................

B 3/4 9-1 3/4.9.6 REFUELING MACHINE.....................................

B 3/4 9-2 3/4.9.7 FUEL HANDLING MACHINE - SPENT FUEL STORAGE BUILDING...

B 3/4 9-2 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION..............

B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE ISOLATION S'CTEM....................

B 3/4 9-3 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and STORAGE POOL.........................................

B 3/4 9-3 3/4.9.12 FUEL HANDLING BUILDING POST-ACCIDENT CLEANUP FILTER SYSTEM..............................................

B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN.......................................

8 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS...........................

B 3/4 10-1 3/4.10.3 R E AC T O R COO LAN T L00 P S.................................

8 3/4 10-1 3/4.10.4 CENTER CEA MISALIGNMENT...............................

B 3/4 10-1 SAN ONOFRE UNIT 3 XII

TABLE 2.2-1 E

REACTOR PROTECTIVE INSTRUNENTATION TRIP SETPOINT LIMITS o

5, 5

FUNCTIONAL LMIT TRIP SETFOINT ALLOWABLE VALUES g

1.

Manuel Reacter Trip Not Applice.ble Not Appilcabie 2.

Lineer Power Level - High -

Four Reacter Coolant Pumps i 110.0% of RATED THERMAL POWER

$ 111.3% of RATED THERMAL POWER Operating 3.

Logarithmic Power Level - High (1) < 0.89% of RATED THERMAL POWER 1 0.96% of RATED THERMAL POWER 4.

Pressurizer Pressure - High 5 2382 psia i 2389 psia 5.

Pressurizer Pressure - Low (2) 1 1806 psia t 1763 psia 3

6.

Contalment Pressure - High 1 2.95 psig i 3.14 psig 7.

Steam Generater % sure - Low (3) > 729 psis 1 711 psia 8.

steam Generater Level - Low

> 25% (4) 1 24.23% (4) 9.

Local Power Density - High (5)

$ 21.0 kw/ft

$ 21.0 kw/ft l

10. Diet - Low 1 1.31 (5)

> 1.31 (5)

11. Reacter Coelant Flow - Low a) su Rate 5 0.22 psid/sec (6)(8) 1 0.231 psid/sec (6)(8) b) Fleer

> 13.2 psid (6)(8)

< 12.1 psid (6)(8)

r c) Step i6.82psid(6)(8) i7.231psid(6)(8) s

'. ?

12. Stcm Gem r?o-Level - High 1 90% (4)

$ 90.74% (4)

.n

{

13. Selsaic - High i 0.48/0.60 f7) 1 0.48/0.60 (7)

14. Loss of Load Turbine stop valve closed Turbine stop valve closed

TABLE 2.2-1 (Continued)

E REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS og TABLE NOTATION 3

(1) Trip may be sawally bypassed above 10-4% of RATED THERMAL POWER; bypass shall be automatically J.

removed when THERMAL POWER is less than or equal to 10-4% of RATED THERMAL POWER.

(2) Value may be decreased manually, to a minimum value of 300 psia, as pressurizer pressure is reduced, provided the margin between the pressurizer pressure and this value is s.aintained at less than or equal to 400 psi; the setpoint shall be increased automatically as pressurizer pressure is increased until the trip setpoint is reached.

Trip may be manuilly bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is greater than or equal to 500 psia.

(3) Value may be decreased manually as steam generator pressure is reduced, provided the margin between the steam generator pressure and this value is maintained at less than or equal to l

200 psi; the setpoint shall be increased automatically as steam generator pressure is increased until the trip setpoint is reached.

~

(4) % of the distance between steam generator upper and low level instrument nozzles.

2 (5) As stored within the Core Protection Calculator (CPC). Calculation of the trip setpoint includes measurement, calculational and processor uncertainties, and dynamic allowances. Trip may be manually bypassed below 10-4% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is greater than or equal to 10-4% of RATED THERMAL POWER.

The approved DNBR limit accounting for use of HID-2 grid is 1.31.

The bypass setpoint may be changed during testing pursuant j

to Special Test Exception 3.10.2.

l 1

(6) DN RATE is the maximum decrease rate of the trip setpoint.

I FLOOR is the minimum value of the trip setpoint.

STEP is the amount by which the trip setpoint is below the input signal unless limited by DN Rate or Floor.

y (7) Acceleration, horizontal / vertical, g.

(8) Setpoint may be altered to disable trip function during testing pursuant to Specification 3.10.3.

M

'5 5

l l

l

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REACT!V!TY CONTROL SYSTEMS SHUT 00WN CEA INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1. 3. 5 All shutdown CEAs shall be withdrawn to greater than or equal to I

145 inches.

APPLICABILITYi' MODES 1 and 2*f.

ACTION:

With a ma=imum of one shutdown CEA withdrawn to less than 145 inches, except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hour either:

Withdraw the CEA to greater than or equal to 145 inches, or a.

b.

Declare the CEA inoperable and apply Specification 3.1.3.1.

SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown CEA shall be determined to be withdrawn to greater than or equal to 145 inches:

a.

Within 15 minutes prior to withdrawal of any CEAs in regulating groups during an approach to reactor criticality, and b.

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

-"See Special Test Exception 3.10.2.

'With K,ff greater than or equal to 1.0.

SAN ONOFRE-UNIT 3 3/4 1-21

~

REACTIVITY CONTROL SYSTEMS REGULATING CEA INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 l

a.

When COLSS is in-service, the regulating CEA groups shall be limited to the withdrawal sequence and to the insertion limits shown on Figure 3.1-2.

The CEA insertion between the Long Term Steady State Insertion Limits and the Transient Insertion Limits is restricted to:

1.

Less than or equal to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval, 2.

Less than or equal to 5 Effective Full Power Days per 30 Effec-tive Full Power Day interval,tand 3.

Less than or equal to 14 Effective Full Power Days per 365 Effective Full Power Day interval.

a.

When COLS$ is out-of service, the regulating CEA groups shall be limited ta the Short Term Steady State Insertion Limit shown on Figure 3.1-2.

The CEA insertion between the Long Term Steady State Insertion Limits and the Short Term Steady State Insertion Limits is restricted to:

1.

Less than or equal to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval, 2.

Less than or equal to 5 Effective Full Power Days per 30 Effec-tive Full Power Day intervai, and

)

3 Less than or equal to 14 Effective Full Power Days per 365 Effective Full Power Day interval.

l APPLICABILITY: MODES 1* and 2*#.

1 ACTION:

l When COLSS is in service and l

a.

With the regulating CEA groups inserted beyond the Transient Insertion Limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, within two hours either:

1.

Restore the regulating CE/. groups to within the limits, or 2.

Reduce THERMAL POWER to less than or equal to that fiaction of RATED THERMAL POWER which is allowed by the CEA group position using the above figure.

See Special Test Exceptions 3.10.2 and 3.10.4.

With K,ff greater than or equal to 1.0.

SAN ONOFRE-UNIT 3 3/4 1-22 AMEN 0 MENT NO. 53

O REACTIVITY CONTROL SYSTEMS ACTION: (Continued) b.

With the regulating CEA groups inserted between the Long Tern Steady State Insertion Limits and the Transient Insertion Limits for inter-vals Steater than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval, operation may proceed provided either:

1.

The Short Ters Steady State Insertion Limits of Figure 3.1-2 are not exceeded, or 2.

Any subsequent increase in THERMAL POWER is restricted to less than or equal to 5% of RATED THERMAL POWER per hour, With the regulating CEA groups inserted be*. ween the Long Tern Steady c.

State Insertion Limits and the Transient Insertion Limits for intervals greater than 5 EFPD per 30 EFPD interval or greater than 14 EFPD per 365 EFPD interval, either:

l 1.

Restore the regulating groups to within the Long Ters Steady State Insertien Limits within two hours, or 2.

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

When COLSS is out of service and the regulatiing CEA groups are inserted beyerid' the Short Tere Steady State Insertion Limit except for surveillance testing pursuant to Specification 4.1.3.1.2, within two hours either:

a.

Restore the regulating LEA group to within the limit, or b.

Reduce thermal power to less than or equal to that fraction of Rated Thermal Power which is allowed by the CEA group position and the Short Ters Stearly State Insertion Limit.

SURVEILLANCE REQUIREMENTS 4.1.',6 The position of each regulating CEA group shall be determined to be within the Transient Insertion Limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the PDIL Auctioneer Alars Circuit is inoperable, then verify the individual CEA positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The accumulated times during which the regulating CEA groups are inserted beyond the Long Tere Steady State Insertion Limits but within the Transient Insertion Limits shall be determined at least once per 24 hourt.

SAN ONOFRE-UNIT 3 3/4 1-23 AMENDMENT NO. 5 3

REACTIVITY CONTROL SYSTEMS i

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l SAN ONOFRE-UNIT 3 3/4 1-24 AMENDMENT NO. 53 I

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REACTIVITY CONTROL SYSTEMS PART LENGTH CEA INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.7 The part length CEA group shall be limited to the insertion limits shown on Figure 3.1-3 with PLCEA insertion between the Long Term Steady State Insertion Limit and the Transient Insertion Limit restricted to:

< 7 Effective Full Power Days per 30 Effective Full Power Day a.

interval,and

)

b.

< 14 Effective Full Power Days per 365 Effective Full Power Day Tnterval.

AFPLICABILITY: MODE I above 20% of RATED THERMAL POWER

ACTION:

With the part length CEA groups inserted beyond the Transient Inser-a.

tion Limit, except for surveillance testing pursuant to specifica-tion 4.1.3.1.2, within two hours either:

1.

Restore th: part length CEA groups to within the limit, or 2.

Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the PLCEA group posi-tion using Figure 3.1-3.

b.

With the part length CEA groups inserted between the Long Term Steacy State Insertion Limit and the Transient Insertion Limit for intervals > 7 EFPD per 30 EFPD interval or > 14 EFPD per 365 EFPD interval, either:

1.

Restore the part length groups to within the Long Ters Steady State Insertion Limit within two hours, or 2.

Reduce THERMAL POWER to less than or equal to 20% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.3.7 The position of the part length CEA group shall be determined to be within the Transient Insertion Limit at letst once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The accumu-lated time during which the part length CEA groups are inserted beyond the Long Tere Steady State Insertion Limit but within the Transient Insertion Limit shall be determined at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

See Special Test Exception 3.10.2.

SAN ON3FRE-UNIT 3 3/4 1 25 AMENDMENT NO. 53

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FIGURE 3.1-3 Part length CEA Insertion Limit Vs. Thermal Power l

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I l

v.g TABLE 3.3-1 h

REACTOR PROTECTIVE INSTRUMENTATION h

MINIMUM e

TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNt:TIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION J

z

[

1.

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

l 2 sets of 2 I set of 2 2 sets of 2 3*, 4*, 5*

7A l

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

j 4.

Pressurizer Pressure - High 4

2 3

1, 2 2#, 3#

5.

Pressurizer Pressure - Low 4

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

w 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#

8.

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

9.

Local Power Density - High 4

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

l

10. DNBR - Low 4

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

l

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#

3*, 4*, 5*

7A g

13. Reactor Trip Breakers 4

2(f) 4 1, 2 5

g 3*,

4*, 5*

7A

14. Core Protection Calculators 4

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

5

15. CEA Calculators 2

1 2(e) 1, 2 6#, 7 l

[

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

3

17. Seismic - High 4

2 3

1, 2 2#, 3#

18. Loss of Load 4

2 3

1(g) 2#, 3#

l TABLE 3.3-1(Continuedj TABLE NOTATION A

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

The provisions of Specification 3.0.4 are not applicable.

(a) Trip may be manually bypassed above 10'4% of RATED THERMAL POWER, bypass shallg%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 shall.g%ofRATEDTHERMALPOWER.e automatically removed when THERMAL POW to 10 During testing pursuant to Special Test Exception 3.10.2 or 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.

(d) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.

(e) See Special Test Exception 3.10.2.

(f) Each channel shall be comprised of two trip breakers; actual trip logic l

shall be one-out-of-two taken twice.

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

ACTION STATEMENTS l

ACTION 1 With the number of channels OPERABLE one less than required by

[

the Minimum Channels OPERABLE requirement, restore the inoper-t able 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 i

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 pro-tective system trip breakers.

I With the number of channels OPERABLE one less than the Total

[

ACTION 2 Number of Channels, STARTVP and/or POWER OPERATION may continue provided the inoperable channel is placed in the bypassed or tripped condition within I hour.

If the inoperable channel is i

bypassed, the desirability of maintaining this channel in the bypassed condition shall be reviewed in accordance with Specifi-cation 6.5.1.6e.

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

SAN ONOFRE - UNIT 3 3/4 3-4 AMENDMENT NO. 53

TABLE 3.3-1 (Continued)

ACTION STATEMENTS With a channel process seasurement circuit that affects multiple functional units inoperable or in test, bypass or trip all associated functional units as listed below:

Process Measurement Circuit Functional Unit Bypassed 1.

Linear Power Linear Power Level - High (Subchannel or Linear)

Local Power Density - High DNBR - Low 2.

Pressurizer Pressure - High Pressurizer Pressure - High Local Power Density - High DNBR - Low 3.

Containment Pressure - High Containment Pressure - High (RPS)

Containment Pressure - High (ESF) 4.

Steam Generator Pressure -

Steam Generator Pressure - Low Low Steam Generator Ap 1 and 2 (EFAS 1 and 2) 5.

Steam Generator Level Steam Generator Level - Low Steac Generator Level - High Steam Generator AP (EFAS) 6.

Core Protection Calculator Local Power Density - High DNBR - Low

.CTION 3 -

With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement, ST ARTUP 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 channel in the tripped condi, tion within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and b.

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

Process Measurement Circuit Functional Unit typassed/

Tripped 1.

Linear Power Linear Power Level - High (Subchannel or Linear)

Local Power Density - High DNOR - Low SAN ONOFRE - UNIT 3 3/4 3 5

\\

TARE 3.3-1(Continued)

A ACTION STATEMENTS 2.

Pressurizer Pressure -

Pressurizer Pressure - High High Local Power Density - High DNBR - Low 3.

Containment Pressure -

Containment Pressure - High (RPS)

High Containment Pressure - High (ESF) 4.

Steam Generator Steam Generator Pressure - Low Pressure - Low Steam Generator AP 1 and 2 (EFAS I and 2) 5.

Steam Generator Level Steam Generator Level - Low Steam Generator Level - High Steam Generator AP (EFAS) 6.

Core Protection Local Power Density - High Calculator DNBR - Low STARTUP and/or POWER OPERATION may continue until tne performance of the next required CHANNEL FUNCTIONil TEST.

Subsequent STARTUP and/or POWER OPERATION may continue if one channel is restored to OPERABLE status and the provisions of ACTION 2 are satisfied.

ACTION 4 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes.

ACTION S With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 6 a.

With one CEAC inoperable, operation may continue for up to 7 days provided that at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , each CEA i

is verified to be within 7 inches (indicated position) of i

all other CEA's in its group.

After 7 days, operation may continue provided that ACTION 6.b is met.*

If the exemption to Specification 3.0.4 is used, Action 6.b aust be met, b.

With both CEACs inoperable, operation may coninue provided that:*

1.

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the DNBR margin reqired by Specification 3 2.4 b (COLSS in service) or Specification 3.2.4.d (COLSS out of service) is satisfied.

Note:

Requirements for CEA position indication given in Technical Specification 3.1.3.2.

SAN ONOFRE - UNIT 3 3/4 3-6 AMEN 0 MENT NO. 53

TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION

_ RESPONSE TIME (SEC) 1.

Manual a.

SIAS Safety Injection Not Applicable Control Room Isolation Not Applicable Containment Isolation (3)

Not Applicable Containment Emergency Cooling Not Applicable b.

CSkS Containment Spray Not Applicable c.

CIAS Containment Isolation Not Applicable d.

NSIS Main Steam Isolation Not Applicable e.

RAS

~~

Containment Sump Recirculation Not Applicable f.

CCAS Containment E'mergency Cooling Not Applicable g.

EFAS Auxiliary Feedwater Not Applicable h.

CRIS Control Room Isolation Not Applicable F

1.

TGIS Toxic Gas Isolation Not Applicable j.

FHIS Fuel Handling Building Isolation Not Applicable k.

CPIS Containment Purge Isolation Not' Applicable i

SAN ONOFRE - UNIT 3 3/4 3-27 l

i

o Table 3.3-5 (continued)

INITIATING SIGNAL AND FUNCTION RESPONSE TIME (SEC) 1 2.

Pressurizer Pressure-Low SIAS (1) Safety Injection (a) HighPressureSafetyInjection 31.2*

(b) Low Pressure Safety Injection 41.2*

(c) Charging Pumps 31.2*

(2) Control Room Isolation Not Applicable (3) Containment Isolation (NOTE 3) 11.2* (NOTE 2)

(4) Containment Spray (Pumps) 25.6*

(5) Containment Emergency Cooling (a) CCW Pumps 31.2*

(b) W Valves (NOTE 4b) 23.2*

(c) Emergency Cooling Fans 21.2*

3.

Containment Pressure-High a.

SIAS (1) Safety Injection (a) High Pressure Safety Injection 41.0*

(b) Low Pressure Safety Injection 41.0*

(2) Control Room Isolation Not Applicable (3) Containment Spray (Pumps) 25.4*

(4) Containment Emergency Cooling (a) CCW Pumps 31.0*

(b) CCW Valves (NOTE 4b) 23.0*

(c) Emergency Cooling Fans 21.0*

b.

CIAS (1) Containment Isolation 10.9* (NOTE 2)

(2) Main Feedwater Backup Isolation 10.9 and Backup Isolation Valves (HV 4048, HV 4052. HV 1105, HV 1106, HV 4047,

.HV 4051)

(3) CCW Valves (Note 4a) 20.9 (4) Mainsteam Isolatior! Valves (HV 8204, 8.9 HV 8205)

(5) Minipurge Isolation Valves 5.9 4.

Containment Pressure - High-High CSAS Containment Spray 23.0*

SAN ONOFRE - UNIT 3 3/4 3-28 AMENDMENT NO. 53

REACTOR COOLANT SYSTEM 3/4.4.10 REACTOR COOLANT GAS VENT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.10 The Reactor Coolant Gas Vent System shall be OPERABLE with:

At least one of valves 3HV0296A or 3HV0296B capable of being powered a.

from an emergency bus and providing a vent path from the reactor vessel head; and, b.

At least one of valves 3HV0297A or 3HV02978 capable of being powered from an emergency bus and providing a vent path from the prc.surizer steam space; and,

. At least one of valves 3HY0298, capable of being powered from an c.

emergency bus and providing a vent path to the containment atmosphere, or 3HV0299, capable of being powered from an emergency bus and pro-viding a vent path to the quench tank; and d.

Valves 3HV0296A, 3HV02963, 3HV0297A, 3HV02978, 3HV0299 and 3HV0298 all closed.

APPLICABILITY:

MODES 1, 2, 3 and 4 ACTION:

a.

With any of valves 3HV0296A, 3HV0296B, 3iN0297A, or 3HY02978 inoperable, operation may continue provided that:

i) power is removed from the inoperable valve (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ;

and,

11) valves 3HV0299 and 3HV0298 are maintained closed and power is removed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and, iii) the inoperable valve (s) is restored to OPERABLE status during the next COLD SHUTDOWN.

b.

With any of valves 3HV0299 or 3HY0298 inoperable, operation say continue provided that:

1) power is removed from the inoperable valve (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; aM,

11) valves 3HV0296A, 3HV02968, 3HV0297A and 3HV02978 are all maintained closed and power is removed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and SAN ONOFRE - UNIT 3 3/4 4-37 AMENDMENT NO.53

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION iii) the inoperable valve (s) is restored to OPERABLE status during the next COLD SHUTOOWN.

c.

The provisions of 3.0.4 are not applicable for entry into MODES 3, 2 l

and 1.

i SURVEILLANCE REQUIREMENTS 4.4.10 Each reactor coolant system vent path shall be demonstrated OPERABLE at least once per 18 months by:

s 1.

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

l 2.

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

t 3.

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

i i

f t

SAN ONOFRE - UNIT 3 3/4 4-38 AMENDMENT NO. 53 l

3/4.5 EMERGENCY CORE COOLING SYSTEMS o

3/4.5.1 SAFETY INJECTION TANKS LIMITING CONDITIOW FOR ODERATION

~ '

3.5.1 Each reactor coolant system safety injection tank shall be OPERABLE with:

a.

The isolation valve open and power to the valve removed, b.

A contained borated water volume of between 1680 and 1807 cubic

feet, c.

Between 1850 and 2800 ppe of boron, and d.

A nitrogen cover-pressure of between 615 and 655 psia.

i APPLICABILITY: MODES 1, 2 and 3.*

ACTION:

With one safety injection tank inoperable, except as a result of a a.

closed isolation valve, restore the inoperable tank to OPERABLF status within one hour or be in at least HOT STANDBY within the next-6 hours and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

I b.

With one safety injection tank inoperable due to the isolation valve being closed, either immediately open the isolation valve or be ir, at least HOT STANDBY within one hour and be in HOT SHU'(DOW within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENT 3 4.5.1 Each safety injection tank shall be demonstrated OPERABLE:

a.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:

1.

Verifying that the contained berated water volume and nitrogen cover pressure in the tanks is within the above limits, and 2.

Verifying that each safety injection tar.k isolation valve is open.

With pressurizer pressure greater than or equal to 715 psia.

SAN ONOFRE - UNIT 3 3/4 5-1 AMENDMENT NO. 53

I EMERGENCY CORE COOLING SYSTEMS i

SURVE!LLANCE REQUIREMENTS (Continued) b.

At least once per 31 days and withirt 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months af ter each solution

volume increase of greater than or equal to IX of tank volume by verifying the boron concentration of the safety injection tank solution.

At least once per 31 days by verifying the fuses removed from each c.

safety injection tank vent valve.

i d.

At least once per 31 days when the RCS presture is above 715 psia, by verifying that the isolation valve operator breakers are padlocked in the open position, At least once per 18 months by verifying that each safety injection e.

tank isolation val've opens automatically under each of the following conditions:

1.

Before an actual or simulated RCS pressure signal exceeds 715 psia, and 2.

Upon receipt of an SIAS test signal.

(

i l

l I

{

I i

l s

l I.

SAN Ch0FRE-UNIT 3 1/4 5-9

3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUT 00W MRGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 ray be suspended for measurement of CEA worth and shutdown margin provided reactivity equivalent to at least the highest estimated CEA worth is available for trip insertion from OPERABLE CEA(s).

APPLICABILITY: MODES 2 and 3a.

ACTION:

With any full length CEA not fully inserted and with less than the a.

above reactivity equivalent available for trip insertion, immedi-ately initiate and continue boration at greater than or equal to 40 gpm of a solution containing greater than or equal to 2350 ppm l

boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored, b.

With all full length CEAs fully inserted and the reactor suberitical by less than the above reactivity equivalent, innediately initiate and continue boration at greater than or equal to 40 gpm of a solution containing greater than or equal to 2350 ppe boron or its --

equivalent until the SHUTDOWN MRGIN required by Specification 3.1.1.

is restored.

SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full length and part length CEA required either partially or fully withdrawn shall be detemined at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

4.10.1.2 Each CEA not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 7 days prior to reducing the SHUTDOWN M RGIN to less than the limits of Specification 3.1.1.1.

"Operation in ICDE 3 shall be limited to 6 consecutive hours.

SAN ONOFRE - UNIT 3 3/4 10-1 AMENDMENT NO. 50

SPECIAL TEST EX,CEPTIONS 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The moderator temperature coefficient, group height, insertion, and power distribution limits of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.1.3.7, 3.2.2, 3.2.3, 3.2.7, footnote C of Table 3.3-1 and footnote 5 of Table 2.2-1, and the Minimum Channels OPERABLE requirement of Functional Unit 15 of Table 3.3-1 may be suspended during the performance of PHYSICS TESTS provided:

a.

The THERMAL POWER is restricted to the test power plateau which shall not exceed 85% of RATED THERMAL POWER, and b.

The limits of Specification 3.2.1 are maintained and determined as specified in Specification 4.10.2.2 below.

APPLICABILITY:

MODES 1 and 2.

ACTION:

With any of the limit.s of Specification 3.2.1 being exceeded while the require-ments of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.1.3.7, 3.2.2, 3.2.3, 3.2.7, footnote C of Table 3.3-1 and footnote 5 of Table 2.2-1, and the Minimum Channels OPERABLE requirement of functional Unit 15 of Table 3.3-1 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 REQUIRENENTS 4.10.2.1 The THERMAL POWER shall be determined at least once per hour during PHYSICS TESTS in which the requirements of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.1.3.7. 3.2.2, 3.2.3, 3.2.7, footnote C of Table 3.3-1 and footnote 5 of Table 2.2-1, or the Minimum Channels OPERABLE requirement of Functional Unit 15 of Table 3.3-1 are suspended and shall be verified to be within the test power plateau.

4.10.2.2 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 Specifications 4.2.1.2 and 3.3.3.2 during PHYSICS TESTS above 5% of RATED THERMAL POWER in which the requirements of Specifications 3.1'.1. 3, 3.1. 3.1, 3.1. 3. 5, 3 L 3. 6, 3.1. 3. 7 3.2.2, 3.2.3, 3.2.7, footnote C of Table 3.3-1 and footnote 5 of Table 2.2-1, or the Minimum Channels OPERABLE requirement of Functional Unit 15 of Table 3.3-1 are suspended.

SAN ONOFRE - UNIT 3 3/4 10-2 AMENDMENT NO.53

REACTlv1TY CONTROL SYSTEM 4 BASES MOVABLE CONTR0C ASSENBLIES (Continued)

The establishment of LSSS and LCOs require that the expected long and short ters behavior of the radial peaking factors be determined.

The long ters behavior relates to the variation of the steady-state radial peaking factors with core burnup and is affected by the amount of CEA insertion assumed, the portion of a burnup cycle over which such insertion is assumed and the expected power level variation throughout the cycle.

The short term behavior relates to transient perturbations to the steady state radial peaks due to radial xenon redistribution.

The magnitudes of such perturbations depend upon the expected use of the CEAs during anticipated power reductions and lead maneuvering.

Analyses are performed based on the expected mode of operation of the NS$$ (base load maneuvering, etc.) and from tnese analyses CEA insertions are determined and a consistent set of radial peaking factors defined.

The Long Term Steady State and Short Tern Insertion Limits are determined based upon the assmed mode of operation used in the analyses and provide a means of preserving the assumptions on CEA insertions used.

The limits specified serve to limit the behavior of the radial peaking factors within the bounds determined froe: analysis.

The actions specified serve to limit the extent of radial menon redistribution effects to those accoanodated' in the analyses.

Ti.e Long and Short Tere Insertion Limits of Specifica-tion 3.1.3.6 are specified for the plant which has been designed for primarily base loaded operation but which has the ability to accomodate a Ifnited amount of load maneuvering.

The Transient Insertion Limits of Specification 3.1.3.6 and the Shutdown CEA Insertion Limits of Specification 3.1.3.5 ensure that 1) the minimum SHUTDOWN MARGIN is maintained, and 2) the potential effects of a CEA ejection accident are limited to acceptable levels.

Long ters operation at the Transient Insertion Limits is not permitted since such operation could have effects on the core power distribution which could invalidate ass eptions used to determine the behavior of the radial peaking factors.

The Part Length CEA Insertion Limits of Specification 3.1.3.7 ensure that adverse power shapes and rapid local power changes which affect radial peaking factors and DNB considerations do not occur as a result of a part length CEA group covering the same axial sepent of the fuel assemblies for an extended period of time during operation.

The CEA fully withdrawn position is defined to be greater than or equal to 145 inches. The extrose limits of CEA travel, fully withdrawn and fully inserted, say be described as the upper electrical limit and lower electrical limit respectively.

i f

l SAN ONOFRE-UNIT 3 8 3/4 1-5 AMEN 0 MENT NO.53

e TA8tE B 3/4.4-1 (Continued)

E Temperature of Minfoum Upper g

Drop Charpy V-Notch Shelf Cv energy Weight

@ 30

@ 50 for Longitudinal 5

Piece No.

Code No.

Noterial Vessel location Results ft - Ib - ft - lb Direction-ft Ib 205-03 C-M31-1 A500CLI Inlet Nozzle Forging 5/E

-20 12.

40 124 5 205-03 C-M31-2 A500CL1 Inlet Nozzle Forging 5/E

-20 12 40 124

-4 205-03 C-M31-3 A500CL1 Inlet Nozzle Forging S/E

-20

-15 50 114 w

205-03 C-M31-4 A500CLI Inlet Nozzle Forging 5/E

-20

-15 50

114 205-07 C-M32-1 A500CL1 Outlet Nozzle Ferging 5/E -20

-20 0

159 205-07 C-H32-2 A500CLI Dutlet Nozzle Forging 5/E -20

-20 0

152 231-01 C-M33-1 A533GABCL1 Closure Head Peel

-40 20 M

M 231-01 C-M24-1 A533GABCLI Closure Head Peel

-30 10 NA NA 231-02 C-M35-1 A533GABCL1 Closure Head Dome

-40 10 NA NA

[

M = Not Available 4

1

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

The OPERABILITY of the Shutdown Cooling System relief valve or an RCS vent opening of greater than 5.6 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs is less than or equal to 285'F.

The Shutdown Cooling System relief valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 100'F above the RCS cold leg temperatures or (2) inadvertent safety injection actuation with two HPSI pumps injecting into a water-solid RCS with full charging capacity and letdown isolated.

3/4.4.9 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant.

These programs are in accurdance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a (g) (6) (1).

Components of the Reactor Coolant System were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition and Addenda through Summer 1975.

3/4.4.10 nCACTOR COOLANT GAS VENT SYSTEM Reactor coolant systein gas vents are provided to exhaust noncondensible gases from the primary system that could inhibit natural circulation core cooling following a non-design bases accident.

The OPERABILITY of at least one reactor coolant system vent path from the reactor vessel head and the pressurizer steam space ensures the capability exists to perform this function.

The design redundancy of the Reactor Coolant Gas Vent Systesi serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, or control system does not prevent isolation of the vent path.

The function, capabilities, and testing requirements of the Reactor Coolant Gas Vent System are consistent with the requirements of Item II.b.1 of NUREG-0737, Clarification of TMI Action Plan Requirements " November 1980.

SAN ONOFRE-UNIT 3 B 3/4 4-10 AMENDMENT NO. $3

' go ore g 8

"'o, UNITED STATES

I NUCLE AR REGULATORY COMMISSION w

s.

I CAsmorow, p. c. 20sss I

%,....+./

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUFPORTING AMENDMENT NO. 64 TO FACILITY OPERATING LICENSE NO. NFF-1C

~

AND AMENDPENT NO. 53 TO FACILITY OPERATING LICENS: NO. NPF-15 SOUTHERN CALIFORNIA ELIS0N COMPANY. ET AL.

SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2 & 3 DOCKET N05. 50 361 AND 50-562

1.0 INTRODUCTION

Southern California Edison Company (SCE), on behalf of itself and the other licensees San Diego Gas and Electric Company, The City of Riverside, California, and The City of Anaheim, California has submitted a number of applications for license amendments for San Onofre Nuclear Generating Station (SONGS), Units 2 and 3.

The NRC staff's evaluation of six of these applications 197, 231, 233, 234, 239 and 155)(referred to as Proposed Change.,

is described below.

Nurbers 2.0 DISCUSSION PCN 197 By letter dated May 12, 1987 the licensees submitted proposed change PCN 197 to the SONGS Units 2 and 3 Technical Specifications. The proposed changes would revise Technical Specification 3.1.3.6, "Regulating CEA Insertion Limits," 3.1.3.7, "Part Length CEA Insertion Limits," and 3.10.2, "Group Height, insertion and Power Distribution Limits," as well as Bases 3/4.1.3, "Novable Control Asserblies."

Technical Specification 3.1.3.6 currently provides restrictions on control elseent assee61y (CEA) insertion limits to periods less than or equal to 14 effective full power days (EFPD) per calendar year. The proposed change would replace 'calencar year" with '365 EFPD interval."

Since calendar year may be interpreted as the period from January through Decester, which is not tie intent of tie specified interval, the proposed change providen greater clarification of the intended interval restriction. The staff finds this nomenclature change acceptable.

The part length CEA insertion limits of Technical Specification 3.1.3.7 are intended to ensure that adverse power shapes and rapid local power changes which affect radial peaking factors and departure fron nucleate boiling (DNB) considerations do not occur as a result of a part length Moesiotz 7pp

i.

2 CEA group covering the san axial segment of the fuel assemblies for an extended period of time during operation.

However, the Specification does not clearly specify the allowable duration within the transiurt insertion limit nor does it clearly address operation within the long term steally ' state insertion limit. The long term steady state lirit is based on the expected variation of the steady state radial peaking factors with burnup.

It serves to limit the behavior of the radial peaking factors within acceptable bounds determined from analyses. The transient insertion limit aids in ensuring that the minimum shutdemn marein is maintained and that the potential effects of a CEA e.iection accident are limited to acceptable levels.

Long term operation at the transient insertion limit is not permitted since such operation could have effects on the core power distribution which could invalidate assumptions used to determine the behavior of the radial peaking factors. The preposed change would restrict the part length CEA grcup insertion to the insertion limits of Figure 3.1 3 with insetLion between the long term steady state insertion limit and the transient insertion limit restricted to intervals less than or equal to 7 EFPD per 30 EFPD interval and intervals less than or equal to 14 EFpD per 365 EFFD interval. This is acceptable since it clarifies the intent of Specification 3.1.3.7 to apply the 7 EFPD per 30 EFPD interval only to insertions beyond the long term steady state insertion limit.

The preposed change would also revise the actions to be taken if part ler.sth CEA groups are inserted beyond the transient insertion limit or between the long term steady state insertion limit and the transient insertion limit for excessive periods of time. The staff finds these revised Action Staterents acceptable.

The applicability of Specification 3.1.3.7 would be revised from Modes 1 and 2 to Mode 1 above 20% of rated power. The staff finds this proposed change acceptable since plant operation is limited to no greater than 5%

of rated power in Mode 2, and below 20% of rated power part length CEA insertion is unrestricted.

In addition the proposed change to Specification 4.1.3.7 would revise the Surveillance Requirement so that part length CEA group positicn is determined every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to be within the transient insertion limit and the accusslated time for insertion beyond the long terin steady state insertion limit but within the transient insertion limit is determined every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The staff considers these surveillance intervals adequate to allow the appropriate proposed actions to be taken in sufficient time. The proposed Surveillance Requirer.ent changes are, therefore, acceptable.

The proposed change would also revise Specification 3.10.2 to allow suspension of the insertion limits of Specification 3.1.3.7 during special physics tests. This is acceptable since the addition of power dependent insertion limits for the part length CEAs r.akes Specification 3.1.3.7 similar to Specification 3.1.3.6 which has previously been included as a special test exception in Specification 3.10.2.

3 Finally, the proposed change would revise the Bases to Specification 3/4.1.3 to clarify the extreme limits of CEA travel (fully withdrawn and fully inserted CEA positions).

The tenns "upper electrical limit" and

lower eltetrical limit" are used to describe the fully withdrawn and fully inserted CEA position.

Furthermore, the withdrawn position would be oefined as greater than or equal to 145 inches. This change is acceptable since it is administrative in nature and provides clarification in terminology which thould avoid misunderstand 1ngs.

PCN.231 By letter dated December 14, 1987 the licensees submitted proposed change PCN-231 which would revise Figure 3.1-2 of Technical Specification 3/4.1.3.6, 'Re assembly (CEA)gulating CEA Insertion Limits" to relax the control eierent powerdependentinsertionlimits(PO!L)atpowerlevelsof 251 of rated therr.41 power or less.

The licensee has requested this change in order to preclude the need to borate prior to startup to assure that the estimated critical CEA position is within the zero power CEA insertion limits.

This will also help reduce the amount of waste water generated during startup.

The proposed change would allow more CEAs to be inserted at power levels below 251 of rated thermal power. This results in higher reactivity insertion rates in the event of a CEA initiated reactivity accident. ThE greater CEA insertion at low power also results in a decrease ir, shutdown margin. Therefore, the licensee reevaluated the three limiting events that are affected by individual CEA worth and required shutdown rargin.

These are the CEA withorawal at low power, the CEA ejection at zero power, and the steam line break at zero power.

The revised FDIL results in a i.aximum reactivity insertion rate of 1.7 x IC'g delts k/k/sec during a CEA withdrawal event at low power compared to 1.1 x 10" delta k/k/sec reported in the Cycle 3 relcad analysis report.

However, for the Cycle 3 reload, the licensee performed a parametric l

study on reactivity insertion rate in order to maximize peak reactor coolantsystempressgre. This resulted in an intermediate insertion rate (less than 1.1 x 10' delta k/k/sec) producing the most adverse results.

The licensee has verified that this remains true even for the requested FD1L revision and, therefore, the previous Cycle 3 analysis for the CEA withdrawal event remains bounding.

The CEA ejection event at zero power is initiated by a higher ejected CEA worth than for Cycle 3 due to the revised PDIL. However, the Cycle 1 reference analysis allowed greater CEA insertion at zero power and, therefore, an even higher ejected CEA worth. The CEA ejection event with the proposed revised FDIL is, therefore, bounded by the reference analysis of Cycle 1.

The most restrictive shutdown margin requirerent is based on the postulated stean line break event at no load operating terperature and the resulting

uncontrolled reactor coolant system cooldewn. The reference analysis (Cycle 1) assured a shutdown margin of 5.15% delta L/k for this event.

The proposed revised PDIL results in a greater calculated shutdc r rargin than 5.151 delta k/k and, thus, is bounded by the reference analysis.

In surrary, although the proposed change allows more CEA insertion at lower power relative to current Cycle 3 allowances, it is still less than was allowed for Cycle 1.

A reevaluation of the limiting events which are adversely affected by greater CEA insertion have shown that the previous reference analysis results are still bounding. The proposed relaxatien of the PDIL at 25% of rated thermal power and below is therefore acceptable.

P C N -2,3,3 By letter dated Decerter 14, 1987 the licensees submitted prorcsed change PCN-233 which would revise Technical Specification 3/4.10.2, 'Groun Height, Insertion and Power Distribution Limits." and Tables 2.2-l'and 3.3-1 for the San OnOfre Nuclear Generating Station (50 HGS) Units 2 and 3.

The first aart of the proposed change would reference footnote 5 of Table 2.21, Meactor protective Instrumentation Trip Setpoint Limits

and footnote C of Table 3.3-1, ' Reactor Protective Instrumentation,' In the body of Special Test Exception 3.10.2.

This part of the change would, also modify (1) footnote 5 of Table 2.2-1 to indicate that the bypass setpoint for core protection calculator (CpC) generated trips my be changed during testing pursuant to Special Test Exception 3.10.2 and(2) footnote C of Table 3.3-1 to indicate that the trip may be e.anually bypassed below 51 of rated thermal power during testing pursuant to Special Test Exception 3.10.2.

The second part of the proposed change would modify Surveillance 4.10.2.2 to reference Specification 4.21.2.

The final change would modify Action 6 of Table 3.31 by inserting an 1

exemption to Specification 3.0.4 in the event of inoperability of one or both control element assembly calculators (CEACs).

CEA bank reactivity worth measurements which are performed during low power physics testing routinely generate abnormal CEA configurations which c4uld cause a CPC trip. Therefore the bistable setpoint for these CPC trips (local power density and Dh6R),must be raised duting physics testing as currently allowed by Special Test Exception 3.10.3.

However, since this specification test exception was developed for partial reactor coolant system flow conditions which are not applicable during normal physics testing, the more appropriate test exception would be 3.10.2.

i Oneoftheproposedchgnges,therefore,allowsresettingoftheplant l

trip bistable from 10~ 1 to 51 power in Special Test Exception 3.10.2 to allow performance of the tests that are now accomplished pursuant to Exception 3.10.3.

This change does not alter the physics test program and is acceptable to the staff.

Another proposed change would reference Surveillance Specification 4.2.1.2 in Surveillance 4.10.2.2.

Since 4.2.1.2 specifies conditions l

within which the linear heat rate is to be deterrined, it is the i

appropriate surveillance to reference. The change, therefore, ensures 1

that the intended surveillance will be performed and is acceptable.

4

5 The final portion of the proposed change would modify Action 6 of Table 3.3-1 by adding a Specification 3.0.4 exerption when one or both CEAts are inoperable. An exemption to Specification 3.0.4 allows plant startup in situations where a specified system or component is in an inoperable condition" Such exemptions are normally authorized in cases where system or coeponent inoperability would be allowed indefinitely under the provis-ions of the Action requirements.

Since operation of the plant with both CEACs inoperable is currently allowed to continue incefinitely under the conditions of Action 6.b. the staff finds the proposed change to be consist-ent with the generic intent of the eAerption and is therefore acceptable.

pCN 234' By letter dated December 14, 1987 the licensees submitted prcposed change FCh-234 which would revise Technical Specification 3/4.5.1, "Safety Injection Tanks". TheexistingLimitingConditionforOperation(LCO) 3.5.1.d requires that each reactor coolant syster safety injection tank be Operable with a nitrogen cover-pressure of between 600 and 625 psig.

This requireeent of the LCO ensures that a sufficient volume of borated water will be istediately forced into the reactor core through the cold legs of the Fenctor Cooiant Syster (RCS) in the event that the RCS pressure falls below the pressure of the safety injection tanks ($1Ts).

This surge of water into the core provides the initial cooling mechanism,

during large pipe ruptures within the reactor coolant pressure boundary.-

The proposed change would revise the required upper limit of the nitrogen cover-pressure from 625 psig to 640 psig. This change would prevert possible violation of the pressure limit in the other SITS due to inleakage from the coesnon fill header when one of the tanks is being filled to m intain its pressure within limits. SCE has evaluated the effect of this higher S!T pressure u of coolant accident (LOCA)pon the previously evaluated large break loss By letters dated April 14 and May 6,19EE SCE prov 4ed the results of tSis evaluation. The higher pressure would cause an additional 160 cubic feet of water to flow out the break but there would still be sufficient water remaining in the $1Ts after blowdown is complete to, fill the downceser. Also, the higher pressure would result in core reflood consancing sooner. The net effect is to reduce the peak fuel clad torperature. The licensee has determined that a 51T stessure in excess of 1000 psig wodTd be necessary to cause an unaccepta>1e amount of flow loss during the blowdown period so that insufficient liquid would remain to fill the downcomer. Therefore, the proposed increase in the upper Itait of the nitropen cover-pressure is acceptable because its effect is bounded by the previous accident analysis.

pounds per square inch gauge (ge would revise the units of pressure from In addition, the proposed chan psig) to pounds per square inch absolute (psia). This will make the units of measurement consistent with other units on the control room pt This change is acceptable because it is administrative in nature anc pr ides consistency witi other pressure ressurerents displayed in th :

mi room, i

6-Finally, the. proposed change would delete the Unit 3 Cycle i specific lower 51T 5. M concentration requirement from the Limiting Condition for Operation.

% change is acceptable because this requirement only applied to4yQ 2 operation of Unit 3, which has been corp'eted.

Therefore, this specification is no Icnger applicable.

FCN-239 By letter dated hover.ber 4,1987 the licensees submitted proposed change PCN-239 which would revise Technical Specification (TS) 3/4.3.2, "Engineered Safety Feature Actuation System Instrumentaticq,* to ct,rrect an editorial discrepancy in Tabla 3.3-5 of the specification. Table 3.3-5, ' Engineered Safety Features Response Times," specifies the Main Steam Isolation Yalve (MSIV) response tie.e in two locations, one associated with the Main Steam isolation Signal (M515)(and the other associtted with the Containment Isolation Actuation Signal CIA!). Amendment 46 for Unit 2 and Arandrent 35 for Unit 3 issued on May 16, 1966, approved an increase in the M51V response time from 6.9 to 8.9 seconds but only ir association with the M515 signal. The proposed change woule increase the M51V response time in association with the CIAS signal from 6.9 seconds to 2.9 seconds to make it consistent with that of the M515 signal. The safety evaluation supporting the M515 signal also applies to the CIAS respense tire. This change will provide consistency in Table 3.3-5 and it will not affect plant equipment or operation nor will it affect previously analyzec accidents. Therefore, this change is acceptaMe.

PCN-155 By letter dated April 27 1984 the Itcensees submitted proposed change pCh-155 which would add $ection 3/4.4.10, Reactor Coolant Gas Vent System" (RCGVS) and its bases to the Technical Specifications. This proposed change was subsequently revised to correct typogr6phical errors by letter dated June 3, 1966.

The primary safety concern with the RCGYS is the ability to isolate the system. The proposed Technical Specifintion would maintain the system isclated and, in the event of a malfunction v?

any valve in the systen, would recove power from the redundant isolation valve (s)inthatventpeth(s). The ACTION, statement provides accepta':le time Itaits for returning the s> stem to operability since the RCGV5 is not used for normal plant operations or for mitigation of any design basis accident. The exception to Section 3.0.4 of the Technical 5pecifications is acceptable for the same reason.

Therefore, the staff finds the proposed Section 3/4.4.10 to be acceptable.

3.0 $UmARY OF STArr EVALUATION The staff has reviewed PCN 197, 231, 233, 234, 239 and 155. We find these changes to be acceptable as proposed and clarified by the licensee's submittals referenced harein.

O 4.0 CONTACT WITH-STATE OFFICIAL The NRC staff has advised the Chief of the Radiological Health Branch, State Department of Health Services, Statt of California, of the proposed determination of no significant hazards consideration. No comments were received.

5.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51.32, an environmental assessment related to PCN-231, 233 and 234 has been published (53 FR 29291) in the Federal Reoister on August 3, 1988. Tne Comission has determined that the issuance of this amendment for those changes will not have a significant effect on the quality of the human enviror. ment.

The remaining changes, PCN-197, 239 and 155 covered by these amendments, involve changes in the installation or use of facility components located within the restricted area. The staff has determined that those portions of the amendments involve no significant increase in the amounts of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupation radiation exposure. The Comission has previously issued proposed findings that the changes involve no significant hazards consideration, and there has been no public coment on such findinge. Accordingly, those portions of the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuantto10CFR51.22(b),noenvironmentalimpactstate-ment or environmental assessment need to be prepared in connection with the issuance of this portion of these ainendments.

6.0 CONCLUSION

Based upon our evaluation of the proposed changes to the San Onofre Units 2 and 3 Technical Specification, we have concluded that:

(1) there is endangered by operation in the proposed manner; and (2)public will not be reasonable assurance that the health and sefety of the such activities will be conducted in compliance with the Comission's regulations and the issuance of the amendments will not be inimical to the comon defense and security or to the health and safety of the public.

Principal Contributors:

L. Kono and D. Hickman Dated: August 3, 1988 l

7590-01

_ UNITED STATES NUCLEAR REGULATORY C01941SS10N SOUTHERN CALIFORNIA EDIS0N COMPANY. ET AL.

DOCKET NOS. 50-361 AND 50-362

~~' ENVIRONMENTAL ASSESSMENT AND FINDING 0F 100 $1GNIFICANT IMPACT Tete United States Nuclear Regulatory Commission (the Commission) is considering issuance of amendments to facility Operating License Nos. NPF 10 l

and NPF-15, issued to Southern California Edison Company, San Diego Gas and Electric Company, The City of Riverside, California and The City of Anaheim, California (the licenteet), for operation of the San Onofre Nuclear Generating l

Station, Units 2 and 3, located in San Diego County, California.

ENVIRONMENTAL ASSESSMENT Identification of Proposed Action:

The proposed amendments would incorporate proaosed changes, identified as PCN-231, 233 and 234, as described below:

Proposed Change PCN-231 is a request to revise Technical Specification 3/4.1.3.6, "Regulating CEA Insertion Limits." The proposed change would revise Figure 3..'-2, relaxing the CEA insertion limits at low power levels to increase operating flexibility and to reduce the volume of radioactive waste water, proposed Change PCN-233 is a request to revise Technical Specification 3/4.10.2, *8roup Neight, Insertion and power Distribution Limits " and Tables 2.2-1 and 3.3-1. The proposed change would modify the existing Technical i

specification to allow use of the correct Special Test Eaception during the i

seasurement of various CEA reactivity worths at low power levels. Tables 2.2-1and3.3-1wouldincorporatethisSpecialTestExceptioninFootnotes(5) i i B & c & + to,t T 5 m 4ep

t-4 l

t I and (C), respectively. Additionally, the proposed change avould specify, in Table 3.3-1, an exemption from the requirements of Specification 3.0.4 if the

~

CEACs are inoperable. PCN-233 would also correr.t a reference error in Surveillance Requirement 4.10.2.2.

4 Proposed Change PCN-234 is a request to revise Technical Specification 3/4.5.1, "Safety Inheticn Tanks." The proposed change would increase the upper limit on SIT cover gas pressure from 625 to 640 psig and change the designated units from psig to psia.

The Need for the Proposed Action:

The proposed changes would provide operational flexibility, clarify and correct the Technical Specifications, and reduce the volume of radioactive waste water.

1 Environmental Impacts of the Proposed Actio 1 The Cossnission has completed its evaluation of the proposed revisions to the Technical Specifications and has concluded that the proposed changes provide reasonable assurance that the facility can be operated safely, The proposed changes do not increase the probability or consequences of accidents, no changes are being made in the types of any effluents that may be released offsite, and there is no significant increase in the allowable individual or cumulative occupational radiation exposure. Changing the insertion limits will decrease the volume of radioactive waste water and changing the $1T cover gas pressure will result in core reflood after a large breat LOCA commencing sooner.

Accordingly, the Commission concludes that this proposed action would result in no significant adverse radiological environmental impact.

With regard to potential nonradiological impacts, the proposed change to the Technical Specifications involves systems located within the restricted

( area as defined in.10 CFR Part 20.

It does not affect nonradiological plant effluents and has no other environmental ispact. Therefore, the Corsnission concludes that there are no significant nonradiological environmental impacts associated with the proposed amendments.

The Notice of Consideration of Issuance of Amendments and Opportunity for Hearing in' connection with this action was pubitshed in the FEDERAL REGISTER on March 31,1988(53FR10452). No request for hearing or petition 4

for leave to intervene was filed following these notices.

i Alternative to the Proposed Action:

l Since the Cossnission concluded that there is no significant adverse environmental effect that would result from the proposed action, alternatives with equal or greater environmental impacts need not be evahated, 4

The principal alternative would be to deny the requested amendments, i

Denial of the request would not reduce environmental impacts of plant operation and in fact would prevent a reduction in the volume of radioactive waste water generated at the facility.

Alternative Ust, of Resources:

This action does not involve the use of resources not previously considered in the Final Environmental Statament Related to the Operation of the San Onofre Nuclear Generating station Units 2 and 3, dated April 1981.

Agencies and Persons Con:v1ted:

The NRC staff has reviewed the licensee's request and did not consult other i

egencies er persons.

FINDING 0F N0 5IGNIFICANT IMPACT The Commission has determined not to prepare an environmental 1spect statement for the proposed license amendment.

I

4 a Based upon this environmental assessment, we conclude that the proposed action will not,.have a significant adverse effect on the quality of the hur.an environment.

For further details with respect to this action, see the application for amendments dated Deces6er 14,1987 and the supplementary information provided by letters dated April 14 and May 6,1988, which.are available for public inspection et the Coevnission's Pubite Document Room,1717 H Street, N. W.,

Washington, is.C., and et the General Library, University of California at Irvine, IrviF California 92713.

Cated at Rockville, Maryland, this 29th day of July,1988.

FOR TH NUCLEAR REGU ORY C0m !SS10N eorge Knighto, Director Projec Directorate Y Division of Reactor Projects - III

!Y, Y and Special Projects Office of Nuclear Reactor Regulation e

a e

-4 r,..

m

et 7590-01 UNITED STATES NUCLEAR #7GULATORY C0KilSS10N

' SOUTHERN CALIFORNIA EDISON COMPANY. ET AL.

DOCKET NOS. 50-361 AND 50-362 NOT' ICE OF ISSUANCE OF AMENDMENTS TO FACILITY OPERATING LICENSES The U.S. Nuclear Regulatory Comunission (Cosmission) has issued Amendment No. 64 to Facility Operating License No. NPF-10 and Amendment No. 53 to Facility Operating License No. NPF-15, issued to Southern California Edison Company, San Diego Gas and Electric Company, The City of Riverside, California and The City of Anaheim, California (the licensees), which revised the Technical Specifications for operation of the San Onofre Nuclear Generating Station, Units 2 and 3, located in San Diego County, California.

The amendment was effective as of the date of issuance.

These amendments revise a nus6er of Technical Specifications (TS), and are in partial response to applications for amendments designated as PCN-231, 233 and 234. The Technical Specifications that are changed by each PCN are as follows: PCN-231 - Figure 3.1-2 of TS 3/4.1.3.6, "Regulating CEA Insertion Limits"; PCN-233 - TS 3/4.10.2, "Group Height, Insertion and Power Distribution Limits," and Tables 2.2 1 and 3.3-1; and PCN-234 - TS 3/4.5.1, "Safety Injection Tanks."

The applications for the amendments comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Cosmissionis rules and regulations in 10 CFR Chapter I, which are set forth in the license amendments, gg%fgTN

1 1

. Notice of Consideration of Issuance of Amendeent and Opportunity for i

Prior Hearing in~ connection with this action was published in the FEDERAL REGISTER on March 31,1988(53FR10452). No request for a hearing or petition for leave to intervene was filed following this notice.

The Comission has prepared an Environmental Assessment related to the action and has determined that an environmental 1spect statement will not be prepared and that issuance of the amendments will not have a significant effect on the quality of the human environment.

For further details with respect to the action see (1) the applications for amendments dated Decester 14, 1907 and supplemental letters dated April 14 and May 6,1988,(2) Amendment No. 64 to License No. NpF-10 and Amendment No. 53 License No. NPF-15, (3) the Comission's related Safety Evaluation and (4) the.

Comission's Environmental Assessment. All of these items are available for I

public inspection at the Cossnission's Public Document Room,1717 H Street N.W.,

and at the General Library, University of California, P.O. Box 19557, Irvine, California 92713. A copy of items (2), (3) and (4) may be obtai ed upon request addressed to the U.S. Nuclear Regulatory Comission, Wapington, D.C.

20555, Attention: Director, Division of Reactor Projects !!!, IV, Y and Special Projects.

Dated at Rockville, Maryland this 3rd day of August 1983.

FOR THE NUCLEAR REGULATORY Co MISSION D. E. Hickman, Project Manager Project Directorate V Division of Reactor Projects - Ill,

!Y, Y and Special Projects Office of Nuclear Reactor Regulation

F" e#

N*

' UNITED STATES NUCtEAR REGULATORY COMMISSION SOUTHERN CALIFORNIA EDISON COMPANY. ET AL DOCKET NOS. 50-361/362 NOTICE OF PARTIAL DENIAL OF AMENDMENTS TO FACILITY OPERATING LICENSES AND OPPORTUNITY FOR HEARING The U.S. Nuclear Regulatory Cossnission (the Comission) has denied part of a request by Southern California Edison Company, San Diego Gas and Electric Company, The City of Riverside, California and the City of Anaheim, California (licensees) for amendments to Facility Operating License Nos. NPF-10 and NPF-15, issued to the licensees, for operation of the Saa Onofre Nuclear Generating

~

Station (SONGS), Units 2 and 3, located in San Diego County, California. The Notice of Consideration of Issuance of Amendments was published in the FEDERAL REGISTER on March 31, 1988 (53 FR 10452).

The amendrents, as proposed by the licensees, would revise Technical Specification (TS) Surveillance Requirement 4.3.3.2(a) by changing the fre-quency of performance of the Incore Detection System Channel Check from 7 days to 31 Effective Full Power Days. The propored change would allow verifi-cation of incere detector operability to be performed in conjunction with other routine surveillances.

The staff finds that th0re are no plant-unique characteristics of SONGS '

or 3 which would justify deviation from the Incore Detection System Operability Requirements specified in the Standard Technical Specifications.

In addition, the staff does not consider the reported error introduced by a

%DN}

r s.

single soft (undetected) detector failure to be insignificant. The degree of compromise to me surement quality which is acceptable is a judgment which should be considered generically.

By Sceptember 9, 1988, the licensees may demand a hearing with respect to the denial described above and any person whose interest may be affected by this proceeding may file a written petition for leave to intervene.

A request for a hearing or petition for leave to intervene must be filed with the Secretary for the Comission, U.S. Nuclear Regulatory Comission, Washington, D.C. 20555, Attention: Docketing and Service Branch, or mcy be delivered to the Comission's Public Document Room,1717 H Street, N.W.,

Washington, D.C., by the above date.

A copy of any petitions should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Comission, Washington, D.C. 20555, and to Charles R. Kocher, Esq., Southern California Edison Company, 2244 Walnut Grove Avenue, P.O. Box 800, Rosemead, California 91770, and Orrick, Herrington &

Sutcliff, Attn: David R. Pigoct, Esq., 600 Montgomery Street, San Frar:isco, California 94111.

For further details with respect to this action, see (1) the application for amendments dated December 14,1987, and (2) the Comission's Notification of Denial forwarded to the licensees by letter dated April 19, 1988, which are available for public inspection at the Comission's Public Document Room, 1717 H Street, N.W., Washington, D.C., and at the General Library, University of California at Irvine, Irvine, California 92713. A copy of Item (2) may be obtained upon request addressed to the U.S. Nuclea, Regulatory Comission, l

I o e e %.

3-Washington, D.C.

20555, Attention: Director, Division of Reactor Projects -

~

III, IV, Y and Special Projects.

Dated at Rockville, Maryland, this 3rd day of August 1988.

FOR THE NUCLEAR REGULATORY COMMISSION 5

Donald E. Hickman, Project Manager Project Directorate V Division of Reactor Projects - 111, IV, Y and Special Projects e

$

ML20205L048

Contents

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1.0 INTRODUCTION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

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ML20205L048

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Attachment:

Attachment:

1.0 INTRODUCTION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

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