Text

Amends 120 & 109 to Licenses NPF-10 & NPF-15,respectively, Revising TS Tables 2.2-1,3.3-1,3.3-3,3.3-4 & Associated Bases
	ML20086M279
Person / Time
Site:	San Onofre
Issue date:	07/14/1995
From:	Fields M; NRC (Affiliation Not Assigned)
To:	;
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ML20086M282	List: ML20086M279; ML20086M295;
References
	NUDOCS 9507240230
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.

pn stag $t UNITED STATES y*.

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NUCLEAR REGULATORY COMMISSION f

WASHINGTON, D.C. 20555 4001

\\ * * * *

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I SOUTHERN CALIFORNIA EDIS0N COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY THE CITY OF RIVERSIDE. CALIFORNIA THE CITY OF ANAHEIM. CALIFORNIA DOCKET NO. 50-361 SAN ON0FRE NUCLEAR GENERATING STATION. UNIT NO. 2 l

AMENDMENT TO FACILITY OPERATING LICENSE l

Amendment No.120 2icense No. NPF-10 1.

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

A.

The application for amendment by Southern California Edison Company, et al. (SCE or the licensee) dated September 15, 1993, as i

supplemented by letter dated September 6, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as l

amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the

(

provisions of the Act, and the rules and regulations of the l

Commission; C.

There is reasonable assurar.ce (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; 0.

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

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

I 9507240230 950714 PDR ADOCK 05000361 P

PDR

. 2.

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

(2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 120, are hereby incorporated in the license.

Southern California Edison Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

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

FOR THE NUCLEAR REGULATORY C MISSION hh h

L' Mel B. Fields, Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

July 14, 1995 l

1 l

l ATTACHMENT T0 LICENSE AMENDMENT

]

AMENDMENT NO.120TO FACILITY OPERATING LICENSE NO. NPF-10

)

DOCKET NO. 50-361 Revise Appendix A Technical Specifications by removing the pages identified j

below and inserting the enclosed pages. The revised pages are identified by Amendment number and contain marginal lines indicating the areas of change.

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

l l

BfnqVf INSERT 2-4 2-4 B 2-3 8 2-3 i

3/4 3-4 3/4 3-4 3/4 3-19 3/4 3-19 3/4 3-26 3/4 3-26 B 3/4 3-la B 3/4 3-la i

l 1

l

E iA8tE 2.2-1 z

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS o

E m

h FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 1.

Manual Reactor Trip Not Applicable Not Applicable

~

2.

Linear Power Level - High -

Four Reactor Coolant Pumps i 110.0% of RATED THERMAL POWER 1 111.0% of RATED THERMAL POWER Operating 3.

Logarithmic Power Level - High (1) 1 0.83% of RATED THERMAL POWER 1 0.93% of RATED THERMAL POWER 4.

Pressurizer Pressure - High 1 2375 psia i 2385 psia 5.

Pressurizer Pressure - Low (2) 1 1740 psia 1 1700 psia 6.

Containment Pressure - High.

5 3.1 psig i 3.4 psig 7.

Steam Generator Pressure - Low (3)

> 741 psia

> 729 psia 8.

Steam Generator Level - Low

> 21% (4)

> 20.::% (4) 9.

Local Power Density - High (5) i 21.0 kw/ft i 21.0 kw/ft

10. DNBR - Low

> 1.31 (5)

11. Reactor Coolant Flow - Low 3E a) DN Rate

< 0.22 psid/sec (6)(8)

< 0.231 psid/sec (6)(8)

E b) Floor I 13.2 psid (6)(8)

I 12.1 psid (6)(8)

]

c) Step i6.82psid(6)(8)

{7.25psid(6)(8) x

[ 12. Steam Generator Level - High i 89% (4) 1 89.7% (4) o

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

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

TABLE 2.2-1 (Continuedl REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS TABLE NOTATION (1) Trip may be manually bypassed above 10% of RATED THERMAL POWER; bypass shall to 10'p% of RATED THERMAL POWER.e automatically removed when THERMAL POW (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 maintained 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.

Trips may be bypassed when pressurizer pressure is < 400 psia.

Bypass shall be automatically removed before pressurizer pressure exceeds 500 psia (the corresponding bistable allowable value is s; 472 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 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.

(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% of RATED THERMAL POWER; bypass shall b automatically removed when THERMAL POWER is greater than or equal to 10'g% of RATED THERMAL POWER.

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

The bypass setpoint may be changed during testing pursuant to Special Test Exception 3.10.2.

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

FLOOR is the minimum value of the trip setpoint.

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

(?) Acceleration, horizontal / vertical, g.

@)

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

)

SAN ON0FRE - UNIT 2 2-4 Amendment No. 64.120

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i

BASES i

Linear Power Level-Hiah l

The Linear Power Level-High trip provides reactor core protection against rapid reactivity excursions which might occur as the result of an ejected CEA, or certain intermediate steam line breaks.

This trip initiates a reactor trip at a linear power level of less than or equal to 111.0% of RATED THERMAL POWER.

Loaarithmic Power Level-Hiah The Logarithmic Power Level - High trip is provided to protect the integrity of fuel cladding and the Reactor Coolant System pressure boundary in the event of an unplanned criticality from a shutdown condition.

A reactor i

trip is initiated by the Logarithmic Power Level - High trip at a THERMAL i

POWER level of less than or equal to 0.93% of RATED THERMAL POWER unless this l

f trip is manually bypassed by the operator. The operator may manually bypass this trip when the THERMAL POWER level is above 10 % of RATED THERMAL POWER; this bypass is automatically removed when the THERMAL POWER level decreases to 10' % of RATED THERMAL POWER.

1 Pressurizer Pressure-Hiah The Pressurizer Pressure-High trip, in conjunction with the pressurizer i

safety valves and main steam safety valves, provides reactor coolant system protection against overpressurization in the event of loss of load without reactor trip. This trip's setpoint is at less than or equal to 2385 psia which is below the nominal lift setting 2500 psia of the pressurizer safety j

valves and its operation avoids the undesirable operation of the pressurizer i

safety valves.

Pressurizer Pressure-Low The Pressurizer Pressure-Low trip is provided to trip the reactor and to assist the Engineered Safety Features System in the event of a loss of Coolant Accident.

During normal operation, this trip's setpoint is set at greater than or equal to 1700 psia.

This trip's setpoint may be manually decreased, to a minimum value of 300 psia, as pressurizer pressure is reduced during plant shutdowns, provided the margin between the pressurizer pressure and this trip's setpoint is maintained at less than or equal to 400 psi; this setpoint increases automatically as pressurizer pressure increases until the trip setpoint is reached.

Trips may be bypassed when pressurizer pressure is

< 400 psia.

Bypass shall be automatically removed before pressurizer pressure exceeds 500 psia (the corresponding bistable allowable value is s 472 psia).

The s 472 psia value represents an allowable value which includes margin to account for instrument loop uncertainties and ensures the 500 psia analytical limit will not be exceeded.

SAN ON0FRE - UNIT 2 B 2-3 Amendment No. 88,120

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 1

BASES Containment Pressure-High The Containment Pressure-High trip provides assurance that a reactor trip is initiated prior to safety injection actuation.

Steam Generator Pressure-Low The Steam Generator Pressure-Low trip provides protection against an excessive rate of heat extraction from the steam generators and subsequent l

cooldown of the reactor coolant. The setpoint is sufficiently below the full i

load operating point of approximately 900 psia so as not to interfere with l

normal operation, but still high enough to provide the required protection in the event of excessively high steam flow. This trip's setpoint may be manually decreased as steam generator pressure is reduced during plant i

shutdowns, provided the margin between the steam generator pressure and this trip's setpoint is maintained at less than or equal to 200 psi; this setpoint increases automatically as steam generator pressure increases until the trip setpoint is reached.

Steam Generator Level-Low The Steam Generator Level-Low trip provides protection against a loss of l

feed =ater flow incident and assures that the design pressure of the Reactor Coolant System will not be exceeded due to loss cf the steam generator heat sink. This specified setpoint provides allowance that there will be sufficient j

water inventcry in the steam generator at the time of the trip to provide a eargin of at least 10 minutes before emergency feedwater is required.

I Local Power Density-High The Local Power Density-High trip is provided to prevent the linear heat l

rate (kw/ft) in the limiting fuel rod in the core from exceeding the fuel

}

design limit in the event of any anticipated operational occurrence. The local power density is calculated in the reactor protective system utilizing l

the following information:

i s.

Nuclear flux power and axial power distribution from the excore flux j

monitoring system, i'

b.

Radial peaking factors from the position measurement for the CEAs; c.

Delta T power from reactor coolant temperatures and coolant flow measurements.

i

~

i l

SAN ONOFRE-UNIT 2 8 2-4 AMENDMENT NO.e*

l

~

gr ty REACTOR PROTEC~Inef INSTRUMENTATION v,

E MINIMUM g

TOTAL NO.

CHANNELS CHANNELS APPLICA8LE g

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE TEM)El ACTION n"

1.

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

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

7A

=

2.

Linear Power Level - High 4

2 3

1, 2 2f,3#

3.

Logarithmic Power Level - High a.

Startup and Operating 4

2(a)(d) 3 1, 2 28,3#

4 2

3 3*, 4*, 5*

7A b.

Shutdown 4

0 2**

3,4,5 4

l 4.

Pressurizer Pressure - High 4

2 3

1, 2 2#,3#

5.

Pressurizer Pressure - Low 4

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

R 6.

Containment Pressure - High 4

2 3

1, 2 2f,3#

7.

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

y 8.

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

id 9.

Local Power Density - High 4

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

10. DNBR - Low 4

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

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

12. Reactor Protection System Logic 4

2 3

1, 2 2f,3#

3*, 4*, 5*

7A

13. Reactor Trip Breakers 4

2(f) 4 1, 2 5

3*, 4*, 5*

7A

14. Core Protection Calculators 4

2(c)(d)(e) 3 1, 2 2f,3f,7

15. CEA Calculators 2

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

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

m g

17. Seismic - High 4

2 3

1, 2 2#,38

18. Loss of Load 4

2 3

1(g) 2f,3#

g.

w 2

u,

TABLE 3.3-1 (Continued)

TABLE NOTATION

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 source range neutron flux monitors may be used in Modes 3, 4, and 5 with the reactor trip circuit breakers open or the Control Element Assembly (CEA)

Drive System not capable of CEA withdrawal.

The provisions of Specification 3.0.4 are not applicable.

(a) Trip may be manually bypassed above 10% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is less than or equal to 10% of RATED THERMAL POWER.

(b)

Trips may be bypassed when pressurizer pressure is < 400 psia.

Bypass shall be automatically removed before pressurizer pressure exceeds 500 psic (the corresponding bistable allowable value is s 472 psia).

(c)

Trip may be manually bypassed below 10% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is greater than or equal to 10% of RATED THERMAL POWER.

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 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 within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the protective system trip breakers.

ACTION 2 With the number of channels OPERABLE one lass than the Total Number of Channels, STARTUP and/or POWER OPERATION may continue provided the inoperable channel is placed in the bypassed or tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

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

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

SAN ON0FRE - UNIT 2 3/4 3-4 Amendment No. 64d46,120

l f

TABLE 3.3-3 (Continued)

TABLE NOTATION l

(a) Trips may be bypassed when pressurizer pressure is < 400 psia. Bypass shall be automatically removed before pressurizer pressure exceeds 500 psia (the corresponding bistable allowable value is s 472 psia).

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

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

l (d) Applicability for SDVS is Modes 1, 2, 3, and 4 when the diesel generator circuit breaker is open.

The provisions of Specification 3.0.3 are not applicable.

The provisions of Specification 3.0.4 are not applicable.

With irradiated fuel in the storage pool.

ACTION STATEMENTS ACTION 8 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

i ACTION 9 - With the number of channels OPERABLE one less than the Total Number i

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

If the inoperable channel is bypassed, i

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

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

1 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 l.

Containment Pressure - High Containment Pressure - High (ESF)

Containment Pressure - High (RPS) 2.

Steam Generator Pressure -

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

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

SAN ON0FRE - UNIT 2 3/4 3-19 Amendment No. 4r148,120

o TABLE 3.3-4 (Continued) j ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES e

E ALLOWABLE Z FUNCTIONAL UNIT TRIP VALUE VALUES w

11. FUEL HANDLING ISOLATION (FHIS) a.

Manual (Trip Buttons)

Not Applicable Not Appilcable b.

Airborne Radiation 1.

Gaseous (8)

(8) w 1

c.

Automatic Actuation Logic Not Applicable Not Appilcable e

i

12. CONTAIMMENT PURGE ISOLATION (CPIS) a.

Manual (Trip Buttons)

Not Applicable Not Applicable b.

Airborne Radiation 1.

Gaseous (6)(7)

(6)(7) 11.

Particulate (6)(7)

(6)(7) 111. Iodine (6)(7)

(6)(7) k c.

Containment Area Radiation (Gasuna)

< 325 mR/hr (MODES 1-4)

< 340 mR/hr (MDOES 1-4) g 32.4mR/hr(MODE 6) 12.5mR/hr(MDOE6) d.

Automatic Actuation Logic Not Applicable Not Applicable h

I

. ~.

TABLE 3.3-4 (Continued)

TABLE NOTATION (1)

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

Trips may be bypassed when pressurizer pressure is < 400 psia.

Bypass shall be automatically removed before pressurizer pressure exceeds i

500 psia (the corresponding bistable allowable value is s 472 psia.

(2)

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

(3)

% of the distance between steam generator upper and lower level instrument nozzles.

(4)

Deleted.

(5)

Actuated equipment only; does not result in CIAS.

(6)

The trip setpoint shall be set sufficiently high to prevent spurious alarms / trips yet sufficiently low to assure an alarm / trip should an i

inadvertent release occur.

(7)

Prior to the completion of DCP 53N, the setpoints for Containment Airborne Radiation Monitor 2RT-7804-1 shall be determined by the ODCM.

(8)

The trip setpoint shall be set sufficiently high to prevent spurious alarm / trips yet sufficiently low to assure an alarm / trip should a fuel handling accident occur.

I

Variable setpoints are for use only during normal, controlled plant heatups and cooldowns.

- Above normal background.

SAN ON0FRE - UNIT 2 3/4 3-26 Amendment No. Mrl48,120

3/4.3 INSTRUMENTATION BASES 3/4.3.1'and 3/4.3.2 REACTOR PROTECTIVE and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the reactor protective and Engineered Safety Features Actuation System instrumentation and bypasses ensure thtt 1) the associated Engineered Safety Features Actuation System action and/or reactor trip will be I

initiated when the parameter monitored by each channel or combination thereof 2

reaches its setpoint, 2) the specified coincidence logic is' maintained, j

3) sufficient redundar.cy is mairteined to permit a channel to be out of service for testing or maintenance, and 4) sufficient-system functional j

capability is available from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the i

assun.ptions used in the accident analyses.

When a protection channel of a given process variable becomes inoperable, the inoperable channel may be placed in bypass until the next Onsite Review Committee meeting at which time the Onsite Review Committee will review and document their judgment concerning prolonged operation in bypass, channel trip, and/or repair. The goal shall be to return the inoperable channel to i

service as soon as practicable but in no case later than during the next COLD SHUTDOWN. This approach to bypass / trip in four channel protection systems is i

consistent with the applicable criteria of IEEE Standards 279, 323, 344 and 384.

l The Core Protection Calculator (CPC) addressable constants are provided f

to allow calibration of the CPC system to more accurate indications of power i

level, RCS flow rate, axial flux shape, radial peaking factors and CEA deviation penalties. Administrative controls on changes and periodic checking

-i of addressable constant values (see also Technical Specifications 3.3.1 and 6.8.1) ensure that inadvertent misloading of addressable constants into the i

CPCs is unlikely.

l The redundancy and design of the Control Element Assembly Calculators (CEAC) provides reactor protection in the event one or both CEAC's becomes in-i operable.

If one CEAC is in test or inoperable, verification of CEAC position is performed at least every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

If the second CEAC fails, the CPC's will use DNBR and LPD penalty factors, which restrict reactor operation to some I

maximum fraction of RATED THERMAL POWER.

If this maximum fraction is exceeded a reactor trip will occur.

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

design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability. The quarterly frequency for the CHANNEL FUNCTIONAL TESTS for these systems is based on the analyses presented in the NRC approved topical report, CEN-327, "RPS/ESFAS Extended Test Interval Evaluation," as supplemented.

The measurement of response time at the specified frequencies provides assurance that the reactor protective and ESF actuation associated with each channel is con.pleted within the time limit assumed in the accident analyses.

l SAN ONOFRE - UNIT 2 B 3/4 3-1 AMENDMENT NO.101 l

3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR PROTECTIVE and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)

No credit was taken in the analyses for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined.

Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or

2) utilizing replacement sensors with certified response times.

The Pressurizer Pressure-Low trips may be bypassed when pressurizer pressure is < 400 psia.

Bypass shall be automatically removed before pressurizer pressure exceeds 500 psia (the corresponding bistable allowable value is s 472 psia).

The s 472 psia value represents an allowable value which includes margin to account for instrument loop uncertainties and ensures the 500 psia analytical limit will not be exceeded.

SAN ON0FRE - UNIT 2 B 3/4 3-la Amendment No. 47,120

p **Eco l',

UNITED STATES ye s

,g NUCLEAR REGULATORY COMMISSION i

't

?#ASHINGTON, D.C. 2%%e001 49.....,o I

SOUTHERN CALIFORNIA EDISON COMPANY i

SAN DIEGO GAS AND ELECTRIC COMPANY THE CITY OF RIVERSIDE. CALIFORNIA IHE CITY OF ANAHEIM. CALIFORNIA DOCKET NO. 50-362 SAN ON0FRE NUCLEAR GENERATING STATION. UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 109 i

License No. NPF-15 1.

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

A.

The application for amendment by Southern California Edison Company, et al. (SCE or the licensee) dated September 15, 1993, as supplemented by letter dated September 6,1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health i

and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

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

E.

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

. 2.

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

(2)

Technical Specifications i

The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 109, are hereby incorporated in the license. Southern California Edison Company shall operate the facility in accordance with the Technical Specifications and the l

Environmental Protection Plan.

l 3.

This license amendment is effective as of its date of issuance to be impleniented within 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION f$W Mel B. Fields, Project Manager l

Project Directorate IV-2 1

Division of Reactor Projects III/IV

[

Office of Nuclear Reactor Regulation l

Attachment:

Changes to the Technical Specifications Date of Issuance:

July 14, 1995 i

i i

i

ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 109 TO FACILITY OPERATING LICENSE NO. NPF-15 DOCKET N0. 50-362 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by Amendment number and contain marginal lines indicating the areas of change.

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

REMOVE INSERT 2-4 2-4 B 2-3 8 2-3 3/4 3-4 3/4 3-4 3/4 3-19 3/4 3-19 3/4 3-26 3/4 3-26 B 3/4 3-la B 3/4 3-la 1

TABLE 2.2-1 p

z REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS o5s r;' FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 1.

Manual Reactor Trip Not Applicable Not Applicable 2.

Linear Power Level - High -

Four Reactor Coolant Pumps 5 110.0% of RATED THERMAL POWER

$ 111.0% of RATED THERMAL POWER Operating 3.

Logarithmic Power Level - High (1) $ 0.83% of RATED THERMAL POWER

$ 0.93% of RATED THERMAL POWER 4.

Pressurizer Pressure - High

$ 2375 psia 5 2385 psia 1

l S.

Pressurizer Pressure - Low (2) 1 1740 psia 1 1700 psia 6.

Containment Pressure - High 5 3.1 psig 1 3.4 psig 7.

Steam Generator Pressure - Low (3) 1 741 psia 1 729 psia 8.

Steam Generator Level - Low 1 21.0% (4) 1 20.0% (4) 9.

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

10. DN8R - Low 1 1.31 (5) t 1.31 (5) 11.

Reactor Coolant Flow - Low 4

k a) DN Rate S 0.22 psid/sec (6)(8)

$ 0.231 psid/sec (6)(8) l g

b) Floor 1 13.2 psid (6)(8) 1 12.1 psid (6)(8)

,T, c) Step i 6.82 psid (6)(8) 5 7.25 psid (6)(8)

I z

[ 12.

Steam Generator Level - High 1.89% (4) 1 89.7% (4) o 13.

Seismic - High i 0.48/0.60 (7) 1 0.48/0.60 (7) i 1

m 14.

Loss of Load Turbine stop valve closed Turbine stop valve closed

i IABLE 2.2-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS TABLE NOTATION j

(1) Trip may be manually bypassed above 10% of RATED THERMAL POWER; bypass l

e automatically removed when THERMAL POWER is less than or equal i

shall p% of RATED THERMAL POWER.

to 10' i

(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 maintained at less than or equal i

to 400 psi; the setpoint shall be increased automatically as pressurizer i

pressure is increased until the trip setpoint is reached. Trips may be j

bypassed when pressurizer pressure is < 400 psia. Bypass shall be i

automatically removed before pressurizer pressure exceeds 500 psia (the corresponding bistable allowable value is s 472 psia).

(3) Value may be decreased manually as steam generator pressure is reduced, provided the margin between the steam generator pressure and this value i'

is maintained at less than or equal to 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.

(5) As stored within the Core Protection Calculator (CPC). Calculation of the trip setpoint includes measurement, calculational and processor l

below 10', ties, and dynamic allowances. Trip may be manually bypassed l

uncertain

% of RATED THERMAL POWER; bypass shall b automatically removed when THERMAL POWER is greater than or equal to 10'g% of RATED THERMAL j

POWER.

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

l The bypass setpoint may be changed during testing pursuant to Special Test Exception 3.10.2.

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

j i

FLOOR is the minimum value of the trip setpoint.

i 111E is the amount by which the trip setpoint is below the input signal unless limited by DN Rate or Floor.

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

i (8) Setpoint may be altered to disable trip function during testing pursuant l

to Specification 3.10.3.

i l

l I

SAN ON0FRE - UNIT 3 2-4 Amendment No. 9,109

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES Linear Power Level-Hiah The Linear Power Level-High trip provides reactor core protection against rapid reactivity excursions which might occur as the result of an ejected CEA, or certain intermediate steam line breaks. This trip initiates a reactor trip at a linear power level of less than or equal to 111.0% of RATED THERMAL POWER.

Loaarithmic Power Level-Hiah The Logarithmic Power Level - High trip is provided to protect the integrity of fuel cladding and the Reactor Coolant System pressure boundary in the event of an unplanned criticality from a shutdown condition. A reactor trip is initiated by the Logarithmic Power Level - High trip at a THERMAL POWER level of less than or equal to 0.93% of RATED THERMAL POWER unless this trip is manually bypassed by the operator. The operator may manually bypass this trip when the THERMAL POWER level is above 10% of RATED THERMAL POWER; thi,s bypass is automatically removed when the THERMAL POWER level decreases to 10' % of RATED THERMAL POWER.

Pressurizer Pressure-Hfah The Pressurizer Pressure-High trip, in conjunction with the pressurizer safety valves and main steam safety valves, provides reactor coolant system protection against overpressurization in the event of loss of load without reactor trip.

This trip's setpoint is at less than or equal to 2385 psia which is below the nominal lift setting 2500 psia of the pressurizer safety valves and its operation avoids the undesirable operation of the pressurizer safety valves.

Pressurizer Pressure-tow The Pressurizer Pressure-Low trip is provided to trip the reactor and to assist the Engineered Safety Features System in the event of a Loss of Coolant Accident. During normal operation, this trip's setpoint is set at greater than or equal to 1700 psia.

This trip's setpoint may be manually decreased, to a minimum value of 300 psia, as pressurizer pressure is reduced during plant shutdowns, provided the margin between the pressurizer pressure and this trip's setpoint is maintained at less than or equal to 400 psi; this setpoint i

increases automatically as pressurizer pressure increases until the trip setpoint is reached.

Trips may be bypassed when pressurizer pressure is

< 400 psia.

Bypass shall be automatically removed before pressurizer pressure exceeds 500 psia (the corresponding bistable allowable value is s 472 psia).

The s 472 psia value represents an allowable value which includes margin to account for instrument loop uncertainties and ensures the 500 psia analytical limit will not be exceeded.

SAN ON0FRE - UNIT 3 B 2-3 Amendment No. M,109

4 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES Containment Pressure-Hioh The Containment Pressure-High trip provides assurance that a reactor trip is initiated prior to safety injection actuation.

l Steam Generator Pressure-Low The Steam Generator Pressure-Low trip provides protection against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the reactor coolant.

The setpoint is sufficiently below the full load operating point of approximately 900 psia so as not to interfere with normal operation, but still high enough to provide the required protection in the event of~ excessively high steam flow.

This trip's setpoint may be manually decreased as steam generator pressure is reduced during plant shutdowns, provided the margin between the steam generator pressure and this trip's setpoint is maintained at less than or equal to 200 psi; this setpoint increases automatically as steam generator pressure increases until the trip setpoint is reached.

Steam Generator Level-Low j

The Steam Generator Level-Low trip provides protection against a loss of feedwater flow incident and assures that the design pressure of the Reactor j

Coolant System will not be exceeded due to loss of the steam generator heat sink. This specified setpoint provides allowance that there will be sufficient water inventory in the steam generator at the time of the trip to provide a margin of at least 10 minutes before emergency feedwater is required.

Local Power Density-High The Local Power Density-High trip is provided to prevent the linear heat rate (kw/ft) in the limiting fuel rod in the core from exceeding the fuel design limit in the event of any anticipated operational occurrence. The local power density is calculated in the reactor protective system utilizing the fol' lowing information:

a.

Nuclear flux power and axial power distribution from the excore flux monitoring system; b.

Radial peaking factors from the position measurement for the CEAs; c.

Delta T power from reactor coolant temperatures and coolant flow measurements.

i SAN ONOFRE - UNIT 3 B 2-4 AMENDMENT NO.'E

' ],

IA!!L' REACTOR PROTECTIVE INSTRUNENTATION

-[

v, MINIMUM E

TOTAL NO.

CHANNELS CHANNELS APPLICA8LE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION i

1.

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

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

7A

[

2.

Linear Power Level - High 4

2 3

1, 2 28.3f 3.

Logarithmic Power Level - High a.

Startup and Operating 4

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

4 2

3 3*, 4*, 5*

7A b.

Shutdown 4

0 2**

3,4,5 4

l 4.

Pressurizer Pressure - High 4

2 3

1, 2 2f,3#

w 5.

Pressurizer Pressure - Low 4

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

1 6.

Containment Pressure - High 4

2 3

1, 2 2f 3#

w 7.

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

4, 8.

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

9.

Local Power Density - High 4

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

10. DNBR - Low 4

2(c)(d)(e) 3 1, 2 2f,38

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 2f,3#

)*,4*,5*

7A

13. Reactor 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#,3f,7

15. CEA Calculators 2

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

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

17. Seismic - High 4

2 3

1, 2 2f,38 m

'i

18. Loss of Load 4

2 3

1(g) 2#,3#

E.

N

TABLE 3.3-1 (Ccntinued)

TABLE NOTATION

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 source range neutron flux monitors may be used in Modes 3, 4, and 5 with the reactor trip circuit breakers open or the Control Element Assembly (CEA)

Drive System not capable of CEA withdrawal.

The provisions of Specification 3.0.4 are not applicable.

(a) Trip may be manually bypassed above 10% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is less than or equal to 10% of RATED THERMAL POWER.

(b) Trips may be bypassed when pressurizer pressure is < 400 psia.

Bypass shall be automatically removed before pressurizer pressure exceeds 500 psia (the correspondim) bistablc allowable value is s 472 psia).

(c) Trip may be manually bypassed below 10% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is greater than or equal to 10% of RATED THERMAL POWER. 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 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 within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the protective system trip breakers.

ACTION 2 -

With the number of channels OPERABuE one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may continue provided the inoperable channel is placed in the bypassed or tripped conditicn within I hour.

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

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

SAN ONOFRE - UNIT 3 3/4 3-4 Amendment No. 53,104, 109

=

TABLE 3.3-3 (Continued)

TABLE NOTATION Trips may be bypassed when pressurizer pressure is < 400 psia.

Bypass (a) shall be automatically removed before pressurizer pressure exceeds 500 psia (the corresponding bistable allowable value is s 472 psia).

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

l r

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

(d) Applicability for SDVS is Modes 1, 2, 3, and 4 when the diesel generator circuit breaker is open.

The provisions of Specification 3.0.3 are not applicable.

The provisions of Specification 3.0.4 are not applicable.

'With irradiated fuel in the storage pool.

ACTION STATEMENTS With the number of OPERABLE channels one less than the Total ACTION 8 -

Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hour:. and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

With the number of channels OPERABLE one less than the Total ACTION 9 -

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

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

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

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. Containment Pressure - High Containment Pressure - High (ESF)

Containment Pressure - High (RPS)

2. Steam Generator Pressure -

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

Low

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

SAN ON0FRE - UNIT 3 3/4 3-19 Amendment No. M7,109 A

TABLE 3.3-3 (Continued)

TABLE NOTATION ACTION 10 -

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

a.

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

b.

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

Process Measurement Circuit Functional Unit Bypassed / Tripped 1.

Containment Pressure Circuit Containment Pressure - High (ESF)

Containment Pressure - High (RPS) 2.

Steam Generator Pressure -

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

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

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

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

ACTION 11 -

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

ACTION 12 -

With the number of OPERABLE channels one less than the Total Number of Channels, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ;-

however, one channel may be bypassed for up to I hour for surveillance testing provided the other channel is OPERABLE.

SAN ONOFRE-UNIT 3 3/4 3-20

{_'..

u E

TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES 5

ALLOWABLE FUNCTIONAL UNIT TRIP VALUE VALUES w 11. FUEL HANDLING ISOLATION (FHIS) a.

Manual (Trip Buttons)

Not Appilcable Not Appilcable b.

Airborne Radiation 1.

Gaseous (B)

(8) l c.

Automatic Actuation logic Not Applicable Not Applicable

12. CONTAIMMENT PURGE ISOLATION (CPIS) w4 a.

Manual (Trip Buttons)

Not App 1(cable Not Applicable b.

Airborne Radiation 1.

Gaseous (6)(7)

(6)(7) 11.

Particulate (6)(7)

(6)(7) 111. Iodine (6)(7)

(6)(7) c.

Containment Area Radiation (Gamma)

< 325 mR/hr (MODES 1-4)

< 340 mR/hr (MODES 1-4) 32.4mR/hr(Mode 6) 32.5mR/hr(MODE 6) k d.

Automatic Actuation Logic Not Appilcable Not Applicable 5

E

l o - -

r TABLE 3.3-4 (Continued)

{

TABLE NOTATION (1)

Value may be decreased manually, to a minimum of greater than or equal to 300 psia, as pressurizer pressure is reduced, provided the margin between the pressurizer and this value is maintained at less than or equal to 400 psia;* the setpoint shall be increased automatically as pressurizer i

pressure is increased until the trip setpoint is reached. Trips may be bypassed when pressurizer pressure is < 400 psia.

Bypass shall be automatically removed before pressurizer pressure exceeds 500 psia (the

[

corresponding bistable allowable value is s 472 psia).

j (2) 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 200 psi;* the setpoint shall be increased automatically as steam generator pressure is increased until the trip setpoint is reached.

(3) % of the distance between steam generator upper and lower level instrument nozzles.

[

(4) Deleted.

l (5) Actuated equipment only; does not result in CIAS.

(6) The trip setpoint shall be set sufficiently high to prevent spurious alarms / trips yet sufficiently low to assure an alarm / trip should an inadvertent release occur.

(7)

Prior to the completion of DCP 53N, the setpoints for Containment Airborne Radiation Monitor 3RT-7804-1 shall be determined by the ODCM.

(8) The trip setpoint shall be set sufficiently high to prevent spurious alarm / trips yet sufficiently low to assure an alarm / trip should a tuel i

handling accident occur.

l

Variable setpoints are for use only during normal, controlled plant heatups and cooldowns.

- Above normal background.

N 1

SAN ONOFRE - UNIT 3 3/4 3-26 Amendment No. 45,107, 109 i

= '.

3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR PROTECTIVE and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the reactor protective and Engineered Safety Features Actuation System instrumentation and bypasses ensure that 1) the associated Engineered Safety Features Actuation System action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, 2) the specified coincidence logic is maintained,

3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4) sufficient system functional capability is available from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions.

The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses.

When a protection channel of a given process variable becomes inoperable, the inoperable channel may be placed in bypass until the next Onsite Review Committee meeting at which time the Onsite Review Committee will review and document their judgment concerning prolonged operation in bypass, channel trip, and/or repair.

The goal shall be to return the inoperable channel to service as soon as practicable but in no case later than during the next COLD SHUTDOWN.

This approach to bypass / trip in four channel protection systems is consistent with the applicable criteria of IEEE Standards 279, 323, 344 and 384.

The Core Protection Calculator (CPC) addressable constants are provided to allow calibration of the CPC system to more accurate indications of power level, RCS flow rate, axial flux shape, radial peaking factors and CEA deviation penalties.

Administrative controls on changes and periodic checking of addressable constant values (see also Technical Specifications 3.3.1 and 6.8.1) ensure that inadvertent misloading of addressable constants into the CPCs is unlikely.

The redundancy and design of the Control Element Assembly Calculators (CEAC) provides reactor protection in the event one or both CEAC's becomes in-operable.

If one CEAC is in test or inoperable, verification of CEAC position is performed at least every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

If the second CEAC fails, the CPC's will use DNBR and LPD penalty factors, which restrict reactor operation to some i

maximum fraction of RATED THERMAL POWER.

If this maximum fraction is exceeded a reactor trip will occur.

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

The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability. The quarterly frequency for the CHANNEL FUNCTIONAL TESTS for these systems is based on the analyses presented in the NRC approved topical report, CEN-327, "RPS/ESFAS 4

Extended Test Interval Evaluation," as supplemented.

The measurement of response time at the specified frequencies provides assurance that the reactor protective and ESF actuation associated with each channel is completed within the time limit assumed in the accident analyses.

SAN ONOFRE - UNIT 3 B 3/4 3-1 AMENDMENT NO.90

=

3/4.3 INSTRUMENTATION BASES 32123 1 and 3/4.3.2 REACTOR PROTECTIVE and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)

No credit was taken in the analyses for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or

2) utilizing replacement sensors with certified response times.

The Pressurizer Pressure-Low trips may be bypassed when pressurizer pressure is < 400 psia.

Bypass shall be automatically removed before pressurizer pressure exceeds 500 psia (the corresponding bistable allowable value is s 472 psia). The s 472 psia value represents an allowable value which includes margin to account for instrument loop uncertainties and ensures the I

500 psia analytical limit will not be exceeded.

l l

SAN ON0FRE - UNIT 3 B 3/4 3-la Amendment No. 36,109 l

___-__ - __ -

ML20086M279

Text

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