Text

Amends 106,98 & 78 to Licenses NPF-41,NPF-51 & NPF-74, Respectively,Revising TS Sections 3/4.1.1.1,6.9.1.9 & 6.9.1.10 to Relocate Shutdown Margin (Reactor Trip Breakers Open) to COLR
	ML17312A743
Person / Time
Site:	Palo Verde
Issue date:	04/30/1996
From:	Thomas C; NRC (Affiliation Not Assigned)
To:	;
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ML17312A744	List: ML17312A743; ML17312A745;
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UNITED STATES NUCLEAR REGULA7ORY COMMISSION WASHINGTON, D.C. 20555-0001 ARI ONA'PU L C S

RV C

CO PANY ET A.

AO' NC OC TN, 50-528.,

N TATION:U T

O.

1'M TO FACI IT OP RAT NG IC NS Amendment No.

106 License No. NPF-41 The Nuclear Regulatory Commission -(the Commission) has found that:

A.

The application for amendment by the Arizona Public Service Company (APS or the licensee) on behalf of itself and the Salt River Project Agricultural Improvement and Power District, El Paso Electric Company, Southern California Edison

Company, Public Service Company of New Mexico, Los Angeles Department of Water and

Power, and Southern California Public Power. Authori.ty..dated.

February 1,

1996, complies with the standards and. requirements of the Atomic Energy Act of-1954, as amended (the Act) and the-Commi'ssion'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 and safety of the public, and (ii)'hat.such activi,ties, 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 2.

E.

The issuance. of.this amendment's in -accordance with '10 CFR'art 51 of the Commission's, regulations and all,appl.icable requirements have been satisfied.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Operating License No. NPF-41 is hereby amended to read as follows:

9'b050.'F0242 9b0430 PDR ADOCK 05000528 P

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The Technical Specifications contained. in Appendix A, as revised through Amendment No.

106, and the Environmental.Protection"Plan contained in Appendix B, are hereby incorporated into this license.

APS shall operate the facility in accordance with the Technical Specifications and the Environmental Protection

Plan, except where otherwise stated in specific license conditions.

3.

This license amendment is effective as of the date of issuance to be implemented within 45 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION harles R. Thomas, 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:

April 30, 1996

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'NO NPF-41

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S 50-58 Replace the following pages of the Appendix A Technical. Speci.fications-with the enclosed pages.

The revised pages are identified by Amendment number and contain marginal lines indi'cating the areas of change.

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

V*

VI 2-5

'8 2-4 3/4 1-1 3/4 3-23 3/4, 3-27 3/4 5-3 3/4 5-4*

3/.4 5-7 8 3/4 3-3 8 3/4 3-4 8 3/4 5-2 6-20a V

VI 2-5 8 2-4 3/4 1-1 3/4 3-23 3/4 3-27 3/4 5-3 3/4 5-4 3/4 5-7 8 3/4 3-3 8 3'/4 3-4 '

3/4 '5-2 6-20a

No changes were made to these

pages, reissued

'to become overleaf.pages.

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0

'f Oi INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION 3/4. 2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT..RATE...........,......""-.-

3/4.2. 2 3/4. 2. 3 3/4.2.4 PLANAR RADIAL PEAKING FACTORS - F..

AZIMUTHAL POWER TILT - Tq...

DNBR MARGINe ~

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3/4.2.5 RCS FLOW RATE..........

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~ ~ ~ ~ -

~ ~ ~ ----- ~ -

~ - ~--

3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE....................

3/4.2.7 AXIAL SHAPE INDEX....................-.....-.--.

~.---'.--

3/4. 2. 8 PRESSURIZER PRESSURE..............

3/4. 3 INSTRUMENTATION PAGE 3/4 2-1 3/4 2"2 3/4 2-3 3/4 2-5 3/4 2-6 3/4 2-7 3/4,2-9 3/4 2-10 3/4. 3. 1 3/4.3. 2 REACTOR PROTECTIVE INSTRUMENTATION.

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.

3/4 3-1 3/4 3-17 3/4. 3. 3 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION.................

INCORE DETECTORS SEISMIC INSTRUMENTATION..............,................

METEOROLOGICAL'NSTRUMENTATION........................

REMOTE SHUTDOWN SYSTEM......................-

~..-.-..

POST-ACCIDENT MONITORING INSTRUMENTATION....'....;....

LOOSE-PART DETECTION INSTRUMENTATION.................

EXPLOSIVE GAS 'MONITORING INSTRUMENTATION....

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND"POWER OPERATION';............

OT STANDBY.......................................

H HOT SHUTDOWN.............

COLD SHUTDOWN -

LOOPS FILLED............................

COLD SHUTDOWN -

LOOPS NOT FILLED........................

3/4 3-37 3/4 3-41 3/4 3-42 3/4 3-45 3/4.3-48

'3/4"3-57 3/4 3-61 3/4 3-63 3/4 4-1 3/4 4-2 3/4 4-.3 3/4 4-5 3/4 4-6 PALO VERDE - UNIT 1 AMENDMENT NO. 27, N, 69

TI G

C D T ON FOR OP~(~ION AND SURVE LANCE RE UIREMI NTS PAGg 3/4.4.2 SAFETY VALVES S HUTDOWN o ~

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0PERATING.:..-;..........................-........,..;......

3/4, 4;-7, 3/4, 4-';8, 3/4.4.3 PRESSURj[ZER PRESSURIZER.....................J.....'.'...'.

AUXILIARYSPRAY.....................................

3/4, 4;-9, 3/4 4-10, 3/4.4.4 STEAM GENERATORS.................... l;... i........,........

3/4 4-11,

,3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE, LEAKAGE DETECTION SYSTEMS.......,l....i........,........

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

3/4 4;-18, 3/4 4;-19, 3/4.4. 6 3/4.4.7 3/4.4.8 CHEMISTRY...,

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SPECIFIC ACTIVITY..........................;;..........

PRESSURE/TEHPERATURE LIMITS 3/4 4;-22, 3/4 4;-25 REACTOR COOLANT SYSTEM..........l.....t........,,.......

PRESSURIZER HEATUP/COOLDOWN LIMITS...l.... *..........

OVERIPRESSURE PROTECl ION SYSTEMS......,...............

3/4 4;-28 3/4 4-31 3/4 4-32 3/4.4.9 STRUCTURAL INTEGRITY.............................

3/4.4.10 REACTOR COOLANT SYSTEH VENTS...........,.......,....,.

II ~ ~ ~

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3/4 4-34 3/4 4-35 4 5 g)Fg g)0~'IG ~Y!~TMS ~CQi 3/4.5.1 3/4.5.2 3/4;5.3 3/4.5.4 SAFETY IN'IECTION TANKS...........................

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~ Il ECCS SUBSYSTEMS OPERATING,.....l..'..~.........,...

ECCS SUBSYSTEMS - SHUTDOWN......l...!..l;...........

REFUELING WATER TANK............L....L..~................

3/4 5-1 3/4 5-3 3/4 5-7 3/4 5-8 PALO VERDE UNI1 1

VI Amendmeit No. R', 106

X441 (G ti d)

V STRU N A ON R

P S TP NT I

TS 0

S (1)

Trip may be manually bypassed above 10 X of RATED THERMAL POWER; bypass shall be automatically removed -when THERMAL POWER is. less. than or equal to 10 'X of RATED THERMAL POWER.

(2)

In MODES 3-4, the value may be decreased manually, to a minimum of 100

psia, as pressurizer pressure is reduced, provided:

(a) the margin between the pressurizer pressure and this value is maintained at less than or equal to 400 psi;. and (b) when the RCS cold leg temperature is greater than-or equal to 485 degrees F,,this value is maintained at least 140 psi greater than the saturation pressure corresponding to the RCS cold leg temperature.

The setpoint shall be increased automatically as pressurizer, pressure is increased until the trip setpoint is reached.

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

(3)

In MODES 3-4, 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)

X of the distance between steam generator upper and lower level wide range instrument nozzles.

(5)

As stored within the Core Protection Calculator (CPC).

Calculation of the trip setpoint includes measurement, calculational and processor uncertainties.

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

PALO VERDE UNIT 1

2-5 Amendment No. 84, i06

I'ABI.E 2. 2-1 (Continued)

REACTOR PIROTECTIVE INSTRUMENTATION'TRIP 'SEYPOINT'IMITS TABLE NOTATIONS (Continued)

(6)

RATE is the <<acimum.rate of decrease of the trip setpoint;.,

There, are no restrictions, on -the rate at,.which..the. setpoint can increase..

FLOOR is the <<inimum value of the trip setpo'int.

RR) is the amount by which the trip setpoi'nt is below the input SigInal

~un ess limitIed by Rate or Floor.

Setpoints are based'n steam generator differentia"I pressure.

(7)

The setpoint <<ay be altered to disable trip function during tesIting I

pursuant te,'Specification 3e10.3.

(8)

RATE is the,<<aximum rate of increase of the trip setpoint.

(The rate at which the setpoint can decrease i's noislower than five. percent per second.)

'CEILING is the <<axiaium value of the trip setpoint.

~N ~is the amount 6y which the trip setpoint, is above the steady state input signal unless li<<ited by the rat%,or, the ceiling.

(9)

Z of the distance between steam generator upper and lower level, narrow, range instrument nor+les.

e

RRMMRQQKM'DRZTZL-ygggggNT, NO. 19

BASES.

REACTOR TRIP SETPOINTS (Continued)

The methodology for the calculation of the PVNGS trip setpoint values, plant protection

system, is discussed-in the CE Document No;"CEN-286(V); Rev.

2, dated August 29, 1986.

Manual Reactor Tri The Manual reactor trip is a redundant channel to the automatic protective instrumentation channels and provides manual'eactor trip capability.

Variable Over ower Tri A reactor trip on Variable Overpower is provided to protect the reactor core during rapid positive reactivity addition excuse.sions.,

This trip function will trip the reactor when the indicated neutron flux power exceeds either a

rate limited setpoint at a great enough rate or reaches a preset ceiling.

The flux signal used is,the average of three linear subchannel flux signals originating in each nuclear instrument safety channel.

These trip setpoints are provided in Table 2.2-1.

Lo ari.thmic Power Level - Hi h 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 unless this trip is manually bypassed by the operator.

The operator may manually bypass this trip when the THERMAL POWER level is above 10-4X of RATED.-THERMAL-POWER;.-this bypass is automatically removed when the THERMAL POWER level decreases to 10-4X of RATED THERMAL'OWER.

Pressurizer Pressure - Hi h 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 below-the. nominal lift,setting of the pressurizer safety valves and its operation, minimizes the undesirable opera-tion of the pressurizer 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 decrease in Reactor Coolant System.inventory and,in the event of an increase in heat PALO VERDE - UNIT 1 B 2"3 AMENDMENT NO. 24

r s removal by the secondary system.

During, normal operation, this trip's setpoint may be manually decreased, to a minIimum value of 100 psia, as pressurizer pressure is reduce,d during plant shutdowns, provided the margin between the pressurizer pressure and Ithis. trip's setpoint -is maintained at less thain or equal to 400 psi; this setpoint increases automatically as pressurizer pressure increases until the",trip setpoint.i.,:reached.-= The",setpoint..must.-also:.be, maintained at least 140 psi greater than, the saturation, pressure corresponding, to the RCS cold le,g temperature whenever the RCS cold leg temperature it eiquhl',

to or greater thais 485 degrees F.

'llhis will ensure safety injection actuation, prior to reactor ves,sel upper head'oid formation in event of RCS depressurization caused by a, steam lline break.

These, are indicated values that, include allowances fior unIcertairity.

The operaitoIr may manually bypass this trip when pressurizer pressure is below 400 psia.

This bypass -is automatically removed -when the pressurizer pressure increases to 500 psia.

in

=ll'lh The Containment 'Pressure - High trip provides assurance that a reactor trip is,initiated in the eve,nt of'ontainment building pressurization due to a pipe break inside the containment, building.

The setpoint for this trip i0 identical to the

. afety i;njection setpoint.

P ~e~r~~.ow The Steam Generator Pressure Low.trip provides, protection in the event of an incr ease in.heat removal by the secondary lsystem and subsequent cooldown of the reactor coolant The setpoint is sufficiently below the full,lo'ad'perating point so as not to interfere with normal operaition, b<st 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,i provii'ded,the margin between the steam generator pIressure 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 normal pressure trip setpoint is reached.

The Steam Generator Level Low trip providles protection against a 'loss of feedwater flow incicient aind assures that the. design pressure"of--the'Reactor i

Coolant System will not be exc;ceded due to a decrease in heat removal by the secondary system.

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 auxiliary feedw@tes is required to prevent degraded core cooling.

oc The, Local Power Density High trip is provided to prevent,,the linear heat rate (kW/ft) in tlie limiting fuel 'rod in the:core.,from exceeding -the fuel design limit in the event of ariy design bases anticipated operational occurr'ence.

The local power density is,caIlcdlated in the, reactor protective system utilizing the f'oliowing information:

PALO VERDE - UNIT. 1 B 2-4 Amendment No. 106

'l M

T L

T TY SYST MS 0

RAKR IMITI ND T 0 FO OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to that specified in the CORE OPERATING LIMITS REPORT.

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With the SHUTDOWN MARGIN less than. that specified in,the CORE OPERATING LIMITS

REPORT, immediately initiate and continue boration at greater than or equal to 26 gpm to reactor coolant system of a solution containing greater than or equal to 4000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SU VEIL ANCE R

UIR MENTS

4. 1.1..1..1 The SHUTDOWN MARGIN..shall.'be determined to be. greater than"or equal to that specified in the CORE OPERATING LIMITS REPORT at least once

per, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of at least the following factors:

l.

2.

3.

4 ~

5.

6.

Reactor.

Coolant System boron concentration, CEA position, Reactor Coolant System average temperature, Fuel burnup based on gross thermal'energy. generation, Xenon concentration, and Samarium concentration.

4. 1. 1. 1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within 2 1.0X delta k/k at least once per 31 Effective Full Power Days (EFPD).

This comparison shall consider at least those factors stated in Specification 4. l.l. 1. 1, above.

The predicted reactivity values shall 'be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a=fuel-burnup of 60:EFPD"after each fuel-loading.

4.1. 1. 1.3 With the reactor trip breakers open** and any CEA(s) fully or partially withdrawn, the SHUTDOWN MARGIN shall be verified within one hour after detection of the withdrawn CEA(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the CEA(s) are withdrawn.

See Special Test Exception 3. 10.9.

- The CEA drive system not capable of CEA withdrawal.

PALO VERDE UNIT 1 3/4 1-1 Amendment No. 23-~

"06

~i

~TRE.

Y Tl' V

~RN - RENT'TIR 'M'P jR:AKERN¹ H T NG CON IT ON FOIR OPEIRATIOI'I 3.1.1.2 a ~

b.

C.

APP CA closed.**

hQLR:

a 0 b.

The SHIJTDOWN,HIARGIN shall be.:gre'.at'er ',th'an.,or equal:.to: that: '.speci fied in the CO~RE OPERATING LIMITS REPORT, and For T<< le. s than or equal to $00';F K., shall be less thah 0.99.

'eactor criticality shall not be. achieved with shutdown group iCEA movement.

PRIE, "...

RN

~INTR l

I R,R'*

Mith the SHUTDOWhl MARGIN less than that specified in the CORE OPERATING LIMITS REPORT, immediately initiate and continue boriation at greater than or equal to 26 gpm to the reactor coolant system of a solution containing greater than or'.equal to-4000-ppm

.boron or equivalent until the required, SHUTDOWN MARGIN is restored, and'ith T << less than or equa'I to 500 F and K., greater than dr equal'o 0.95, 1immediiately vary CIEA positioiss anif/or initiate and continue boration,at greater than or equal to 26 gpm to the reactor coolant system of a,solution containing greater than or equal to 4000 ppm boron or equivalent -until the rei]uired K, is restored.

~NT IE N

TRT 4.1.1.2.1 With the reactor trip bre-kers closed~*,

the, SHUTDOWN HARGIN shall, be determined to be greater than or equal to that specif'ied in the CORE OPERATING LIHITS REPORT:

a.

Within I hour'fter detection of an inoperable CEA(s) and at least; once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the CEA(s) is inoperable.

See Special Test Exceptions 3.10.1 and 3.10.9

- The CEA drive system capable of CEA withdrawal.

PALO VERDE -'NIT 1 3/4 1-2 Amendment No. 83.;69,98

TA

. -3 ont'n ed

~AOTAT ON (a)

In HODES 3-4, the value may be decreased manually, to a minimum of 100

psia, as pressurizer pressure is reduced, provided:

(i) the margin between the pressurizer pressure and this value is

'aintained at less than or equal to 400 psi; and (ii) when the RCS cold leg temperature is greater than or equal to 485 degress F, this value is maintained at least 140 psi greater than the saturation pressure corresponding to the RCS cold leg temperature.

The setpoint shall be increased automatically as pressurizer pressure is increased until the trip setpoint is reached.

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

(b)

In HODES 3-4, the 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.

(c)

Four channels

provided, arranged in a selective two-out-of-four configuration (i.e., one-out-of-two taken twice).

(d)

The proper two-out-of-four combination.

(e)

Input to channels.

The provisions of Specification 3.0.4 are not applicable.

~CNT ACTION 12

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

ACTION 13-With the number of channels OPERABLE one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may continue provided the inoperable channel is.placed in the bypassed or tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . If the inoperable channel is

bypassed, the desirability of maintaining this channel in the bypassed condition shall be reviewed in accordance with Specification 6.5. 1.6.g.

The channel shall be returned 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 1.

Steam Generator Pressure

Steam Generator Pressure Low Low Steam Generator Level 1-Low (ESF)

Steam Generator Level 2-Low (ESF) 2.

Steam Generator Level Steam Generator Level Low (RPS)

(Wide Range)

Steam Generator Level 1-Low (ESF)

Steam Generator Level 2-Low (ESF)

PALO VERDE - UNIT 1 3/4 3-23 Amendment No. 87-;96, 106

ACTION,14-ACTION. 15-ACTION 16-ACTION 17 ACTION 18

-'ABLE 3.'3-3 (Caint'inued,'I

-ACTION: STATEMENTS (ContinL!ed) hWith the numbei ofchannels

.OPFRABLE one less than the, Mini~urn Channels OPERABLE,.STARTUP and/or POWER OPERATION.may continuie provided the following cohdi'ti'dns ares'atisfied:

. a.

'Ver:ify that..one. of: the: inoperable channels has"been bypassed and.place the. other-i'noperab'le channel in tl'ie trippjed I

conIdition withir! I,hour.

hb.,

All functional units affected by, the bypassed/tri'ppkd

'hannel shall also. be: p,'laced in the bypassed/tripped condition as Usted bel(tw:

f'.rbcess Measure!!lent-Circuit Functional'nit, Bypassed/Tripped 1.

steam Generator 'P'ressur;e Steam Genei ator Pre'ssi'>re' Low I,.ow Steam.Generator hL'evel 1 - Low=(ESF)

Steam hGeneratoi Level 2 -

Low (ESF) 2;,

-Steam Generator..'Level

- Low Steam,Gene'rator Leviel: -.Low,(RPS)

(Wide Rahnge),Steam:Generator Le'viel 1 -.Low (ESF)

Steam Generator Level 2 -

Low (ESF)

STARTU'P and/or.

POWER COOPERATION r~ayh continue iuntilt the per'fora'nce, of the next required CHANNEL FUNCTIOhlAL TEST,.

Subsequent STARTUP and/or PiDWER..OPERATION..may".continue,hif...one. channel is restored to OPERABLE statu0 a'nd'th'e provisions of ACTION 13 are satIisfied.

With the number of OPFRABLE charinel's one. less than the Total Number of Clharine'ls,, r'estork -t'e'noperable channel to OPERABLE status within 48.,hours or be in at least'HOT STANDBY within 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,s acri ini HOT SHUTDOWN wittiiin the fol,lowirig.h6..hours.

With tlhe,numbe'r.of OPERABLE.clhanne'Is one.les!

than the ITotall Numtier of Channe'Is, be in at, least HOT-.STANDBY withi'n 6i hour;s and in.at,.least HOT SHUTDOWN within the following 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s>

howe.ver, o'e channel may. be bypassIBd for up.to,l hour for.',,',

survei'llanci!,testing provided the other.channel is O'PERABLE.

-Withi:the number 'of OPERABLE clhanne'Is'one 'less'han the IMihieIum Number of Channels; restore the,'inoperable channel to dPERASLEI status:,within 48 'hours..or b'e -in at 'least." HOT.STANDBY wi'thin thIB

.next,6.,hours,Land-in. COLO SHUTDOWN within. the fol1ohwing I30I hcIurs.

With the, number of,,OPERABLE chanriels on!!.Ress than the IMiIii!!Ium Number of Ch'anne1ls; operition Ay Wontihue'or up to'6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; After 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months operation maji cbntinue provided, at least 1 trairii of,essentia'I filtration is,in opieratione otherwhise, be in HOT STANDBY'itl>in the next.6: hours and in tCOLD. SHUTDOWN with'in 'the foll,ow'ing-30 hour.s.

h PALO VERDE - UNIT 1 3/4 3-24 AMENDMENT NO 27

TABLE 3.3-4'.(Continued)

TABLE "NOTATIONS (1)

In MODES 3-4, the value may be'>> decreased manually, to a minimum of 100

psia, as pressurizer pressure is reduced, provided:

(a) the margin between 'the.pressurizer presure and this value is maintained at less than or equal-to 400 psi.;

and (b) when the RCS cold 1'eg temperature is greater than or equal to 485 degrees F, this value 'is maintained, at least 140 psi greater than the saturation pressure corresponding to the RCS cold leg temperature.

The setpoint shall be increased automatically as pressurizer pressure is increased until the trip setpoint is reached.

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

(2)

X of the distance between steam generator upper and lower level narrow range instrument nozzles.

(3)

In NODES 3-4, 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)

X of the distance between steam generator upper and lower level wide range instrument nozzles.

PALO VERDE UNIT 1 3/4'-:27 Amendment No. 408-,106

4~

il(

0 SYST MS UBSYST

P LIMIT G

CO TION FOR OP RATION 3.5.2 Two independent Emergency Core.Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:

a.

One OPERABLE high-pressure safety inject'ion pump, b.

Co One OPERABLE low-pressure safety injection pump, and An independent OPERABLE flow, path capable of taking suction from the refueling water tank on a safety injection actuation signal and automatically transferring suction to the containment sump on a

recirculation actuation signal.

MODES 1, 2, and 3*.

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

b.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within,90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

The current value of the usage factor for each affected injection nozzle shall be provided in,.this Special Report whenever its value exceeds 0.70.

With pressurizer pressure greater than or equal to 1837 psia, or RCS cold leg temperature greater than or equal to 485 degrees F.

PALO VERDE - UNIT 1 3/4 5-3 Amendment No.iO6

ill EMERGENCY CORE COO~ING,'SYSTEMS

~EE 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a.

At least once per"12 Ihours by. verifying th'at the"f'ol.lowing"valves are in the indicated positions with.the. valves key-locked, shut:

~elve oraber Va'Ive Function

/alee Position 1.

SIA HV-604 1.

H01 LEG INilECTIUN 1.

SAUT 2.

SIC HV-32!1 2.

IHOT LFG INilECTION 2.

SHUT 3.

SIB HV-609 3.

HOT LEG INJECTION.

3.

SHUT 4.

SID HV-331 4.

IHOT LEG INJIECTION.

SHUT.

b.

At least once per 31 days by:

1.

Verify'ing that eaIch. valve (manual, pov(er-operated, or automatic)i in the flow path that is not locked;

sealed, or otherwise secured in position,, is in its correct, pbsiti6n, and C.

2.

Verifying that the ECCS piping is full of water by vehting, the accessible discharge piping high. points.

By a visual inspection which verifies that no loose debris (rags,'rash, clothing, etc.) is present in the'ont0inment which could be transported to thIe contaiinmIent sump and cause restriction of thd pUmp'uctions during LOCjh conditions.

This visual inspect'ion shall be performed:

1.

For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRI1'Y, and 2.

At least once daily of the affected areas withIin containment by containment entry and during tlhe final entry when CONTAINMENT INTEIGRITY is e. tablished.

d.

At least onc: e per 18 months.by:

PALO VERDE UNIT 1

3/4 5-4 AMENDMENT NO, 79

.5 SS

-SUTO M T G

0 ON FO OP RATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a.

An OPERABLE high pressure safety injection pump, and b.

An OPERABLE flow path capable of taking suction from the refueling water tank on a safety injection actuation signal and automatically transferring suction to the containment sump on a recirculation actuation signal.

NODES 3* AND 4.

@~0 a ~

b.

With no ECCS subsystem OPERABLE, restore at least one ECCS subsystem to OPERABLE status within I hour or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

In the event the ECCS is actuated and injects water, into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9;2'ithin '90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

SURVEILLA C RE UIRENENTS 4.5.3 The ECCS subsystem shall be demonstrated OPERABLE per the applicable surveillance requirements of Specification 4.5.2.

With pressurizer pressure less than 1837 psia and RCS cold leg temperature less than 485 degrees F.

PALO VERDE UNIT I 3/4 5-7 Amendment No. i06

0 ill p

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 wi.th certified response times.

During normal operation, the low pressurizer pressure trip setpoint may be manually decreased, to a minimum value of 100 psi'a, 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.

This setpoint must also be maintained at least 140 psi greater than the saturation pressure corresponding to the RCS cold leg temperature whenever the RCS cold leg temperature is equal. to. or greater than 485 degrees F.

This will ensure safety injection actuation prior to reactor vessel upper head void formation in event of RCS depressurization caused by a steam line break.

These are indicated values that include allowances for,,uncertainty.

The operator may manually bypass the low pressurizer pressure trip when pressurizer pressure is below 400 psia.

This bypass is automatically removed when the pressurizer pressure increases to 500 psia.

3 4 N

NATO 4

T 0 N

OR NG STRUM NTATION The OPERABILITY of the radiation monitoring channels ensures that:

(1) the radiation levels are continually, measured in the areas served'y the individual channels and (2) the alarm or automatic action is initiated when.

the radiation level trip setpoint is exceeded.

4.

CO T

TO The OPERABILITY of the incore detectors with the specified minimum comple-ment of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the-reactor core.

ON The OPERABILITY of the. seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety.

This capabil,ity is required to permit "comparison of the measured response to that used in the design basis for the facility to determine if plant shutdown is required pursuant to Appendix A of 10 CFR Part 100.

.The instrumentation is consistent with the recommendations of Regulatory Guide 1. 12,. "Instrumentati'on for Earthquakes,'"

April 1974 as identified in the PVNGS FSAR.

The seismic. instrumentation for the site is located in Table 3.3-7.

3 4.

0 S

MNTTO The OPERABILITY of the meteorological i'nstrumentation ensures that suffi-cient meteorological data are available for estimating potential, radiation doses to the public as a result of routine or accidental, release of radioactive materials to the atmosphere.

This capability is required-to evaluate the need for initiating protective measures to protect= the health and:safety of the public and is consistent with the recommenda+ions of Regulatory Guide 1.23 "Onsite Meteorological Programs,."

February 1972.

Wind speeds less than 0.6 MPH cannot be measured by the meteorologi'cal instrumentation.

1 PALO VERDE - UNIT I B 3/4 3-3 Amendment No. 87,106

KIKHH MIBi The OPERABIL'ITY'. o'IF the'emote shutdowni system ensures that suffici'ent

'capability is available to-permit. safe shutdown-and-maintenance"of--HOT STANDBY of the facility from locations outside;of,the conttol room.

This; capabil'ity

.is;,required.,in. the.,event..control room,,habi.tabi:.l,ity, is:.lost,,and;.is..consistent.

with General Design Criter'ion 19 of.10 CFR Part i50i, The, parameters selected to be monitored ensure that'1) the condition of the reactor is known, (2) conditiions in the RCS are 'known,.(3). the ste'am generators are'vailable for residual heat removal (4) a source of. water, is available for makeup to the

RCS, and (5) tHe-cha~rging'yste'm is avai1able"to-makeup water to the RCS.

The OPERABILITY o)F the remote shutdown s~'t'st'em insures that,a fire will not preclude achieving-sa'fe shutdown.

The 'remot'e shutdown

system, instrumentation, control and power circuits: and disconnect switches necessary

~ to eliminate effects of the fire and allow operation of.instrumentation, control and power circuits required to achieve and,maintain a safe, shutdown.

condition are independent of areas where a fire could damage systems riormally used to shutdown the reactor.

-This capability is consistent with General

,Design Criterion 3 and Appendix R to 10 CFR 50.

The alternate disconnect methods.or -power'r control'icircuits ensure that sufficient capability'is availablle to permit shutdown andmaintenance-.;of col'd shutdown: of'h'e faci;lity by 'relying on additional'perator actions at:- local control stations rather than at the RSP.

CCIQKIK EUIIIOJlUIL IILIaaE The OPERABILITY of the post-'-accident monitoring instrumentation ensures that sufficient information is available one selected pl.int pairame'ters'to

'onitor and assess these variables followilIig an accident.

This capabil,ity.is consistent with the recommendations of Regulatory'uide"1:97; "InstrumeInthtion for Light-Mater-Cooled Nuclear P'lants to Assess Plant Conditions During and

,'Following an Accident," December 1975 and NURIEG 0578, "TMI-2 Lessons Ldarhed Task Force Status Report and',Short-.Term,. RecomtIendations,."

The containment high range area monitors (RU-148 8 RU-149) arid -the main steamline radiation'monitors.

(RU-139 -AEB and 'RU~140 -ALB), are; in'able--3 3I-'6J

'The, high. range 'effluent.lt>onitors.and

.amplers',

(RU-.,144 and RU-'-146) at e-in. the,

-ODCH.

The. containment hydrogen monitors are in iSpecification 3'/4;6;-5.1..

Thee Post"Accident SamplIing System (RCS coolant)

'iIs i'n '~lable 3.3-'.

The Subcooled Margin Monitor. (SMH), tke Reit Qunction,.Thermocouplh

'(HJTC),

and. the Core Exit Theieocouples (CET) comprise the Inadequate Core Cooling (ICC) in'strument <tion required by Item II.F.2"NUREG-0737, the Post THI-2 Action Plan.

The functii'on '.of'he ICC i'nstrumeritation. 'is, to enhahce the

,abil.ity of the plant operator to diagnose the, approach.-'to. existance of,,and recovery from ICC.

Additionallly', they.aid in 'tracking

.'re* actor cool.ant.

inventory; Thes'e.instruroents're included in the,,TechnicaIl: Specifications at the request of'NRC Generic Letter 83-37.

These ar'e not required by thj accident analysis,,nor td bring the plant to,Cold Shutdown.

PALO VERDE.- UNIT 1 B'/4. 3-4 Amendment No. Sk,'106

3/4. 5 EMERGENCY CORE 'COOLING SYSTEMS ECCS BASES

".~

i'/4.

5. 1 SAFETY INJECTION TANKS The OPERABILITY of each of the Safety Injection System (SIS) safety injection tanks ensures that a sufficient volume..of borated.'water. will be immediately forced into the reactor -core -through each of the cold legs in the event the RCS pressure falls below the pressure of. the safety injection tanks.

This initial surge of water into the RCS provides the initial cooling mechanism during large RCS pipe ruptures.

The limits on safety injection tank volume, boron concentration, and pressure ensure that the safety injection tanks will adequately perform their function in the event of a LOCA in MODE 1, 2, 3, or 4.

A minimum of 25X narrow range corresponding, to 1790 cubic feet and a maximum of 75K narrow range corresponding to 1927 cubic feet of borated'ater are used in the safety analysis as the volume in the SITs.

To allow for instrument accuracy, 28X narrow range corresponding to 1802 cubic feet and 72X narrow range corresponding to 1914 cubic feet, are specified in,the Technical Specification.

A minimum of 593 psig and a maximum, pressure of 632 psig, are used in the safety analysis.

To allow for instrument accuracy 600 psig minimum and 625 psig maximum are specified in the Technical Specification.

A boron concentration of 2000'.ppm.minimum. and"4400 ppm maximum are used in the safety analysis.

The Technical. Specification lower limit.of.-2300: ppm in the SIT assures that the backleakage from RCS will not dilute the SITs below the 2000 ppm limit assumed in the safety analysis prior to the time when draining of the SIT is necessary.

The SIT isolation valves are not single failure proof; therefore, whenever the valves are open power shall be removed from these valves and the switch keylocked open.

These precautions ensure that, the,.SITs, are, avai,lable during a Limiting Fault.

The SIT nitrogen vent valves are not single failure proof-'against depressurizing the SITs by spurious opening.

Therefore, power to the valves is removed while they are closed to ensure. the safety analysis assumption of four pressurized SITs.

All of the SIT nitrogen vent valves are required to be operable so that, given a single failure, all four SITs may still be vented during post-LOCA long-term cooling.

Venting the SITs provi'des for -SIT depressurization capability which ensures the timely establishment of shutdown cooling entry conditions as assumed by the,,safety.analysis for smal.l. break LOCAs..

The limits for operation with a safety injection tank inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring, concurrent with failure of an additional safety injection tank which may result in unacceptable peak cladding tempera-tures.

If a closed isolation valve cannot be immediately opened, the full capability of one safety injection tank is not available and:prompt action is required to place the reactor in; a: MODE where this capability is not required.

For MODES 3 and 4 operation with pressurizer pressure less than 1837 psia the Technical Specifications require a minimum of 57X wide range corresponding PALO VERDE - UNIT 1 B 3/4 5-1 AMENDMENT NO.

28

B S

JLLI

@9'Continued) to 1361 cubic feet and a maximum of 75X narrow range corresponding to 1927 cubic feet of borated water per tank, when three, safety,injection tanks are operable and a minimum of 36X widie range corresponding tio 908 cubic feet and a

maximum of 75X narrow range corresponding to 1927,cubic feet per tank, when four safety injection tanks are olperab'le at a minimum pressurize of 235 psig and a maximum pressure of.625 psig.

To allow for in. trument inaccuracy, 60%%d wide range instrument corresponding to 1415 -cubic feet,. and.72X. narrow range instrument corresponding to 1914 cubic feet', when"three safety"injection tanks are operable, and 39X wide range instrument'corresponding to 962 cubic feet, and 72X narrow range instrument correspondipg ltol19)4,cubic feet, when f'our SITs are operable, are specified in the Technical, Slpecifications.

To allow for instrument inaccuracy 254 psig is spiecified in the Technical Specificatioris.

The instr umentation vs.

volume correlation for the SITs is as follows:

3hhuae

~erg~~a l

l g g!~~n 962 ft;,

<OX 39X 1415 ft;,

<OX 60X 1802 ft,,

28X 78X 1914 ft 72X 83X

.MME.

"IR'5-i The OPERABILITY of'wo separate and independen't ECCS subsystems with the indicated RCS pres, sure greater than or equal to 1837 psia,-or with the.

indicated RCS coldl leg temperature greater than. or equal to,.485 degrees, F,

ensures 'that, sufficient. emergency ccire cool.ing capability, wil'1'e available in, the event of a LOCA assum'ing the loss of 'one s'ubCys'tern through any sing'le failure consideration.

These

'indicated valhesi include allowances for uncertainties.

Either subsystem operating in,conjunct,ion with the safety injection tanks is capable of supplying sufficient core cooling to limit the, peak cladding temperature.s within acceptable limits for all postulated'Ibreak sizes ranging from the double-ended break of the largest RCS. cold leg pipe downward.

In addi,tion, each 'ECCS subsystem prtovtides long-term-core-coo'ling capability in the recirculation mode duringi the accident'recovery period.

i The Mode 3 safety analysis credits one HPSI pump to provide negati've'eactivity insertion to protect the core and RCS, following a steam line'r'eak

.when RCS cold leg temperature is 485 degrees F or greater.

Requiring two operable ECCS subsystems in the situation will ensure one HPSI pump is I

available assuming single failure of the other HPSI pumps.

Mith the RCS cold leg temperature below 485'. degrees..F,.

one OPERABL'E ECCS subsystem is acceptable without single failure con.ideration on the; basis of the stable reactivity condition"of the reactor.

and the. limited. core cooling requirements.

The trisodium phosphate dodecahydrate (TSP) stored in dissolving basket's located in the containment basement is provided to minimize the possibility,of, corrosion crackinlg of certa.in metal component.s. during operation of the,ECCS, following a LOCA.

1'he TSP, provided this protection by dissolving in the sump, water and causing its final pH to be raised tio.greater than or equal to 7.0i The surveillance requirements provideci tio ensure OPERABILITY of each'omponent ensure that at a minimum, the assumption.s used in 'the safetylanlalysep are met and that subsystem OPERABILITY is maintainied.

Surveillance requirei ments for throttle valve position stops ancl flow balance testing provide PALO VERDE - UNIT 1 B 3/4 5-2 Amendment No. iO6

ADM S

RAT CONTROLS MI S PORT 6.9.1.9 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any rema'ining. part of a

-reload "cycle for the fol.lowing:.

'a ~

b.

C.

d.

e.f.

g, h.i.

J

~

k.l.

Shutdown Margin Reactor Trip Breakers Open for Specification 3.1.1.1 Shutdown Hargin - Reactor Trip Breakers Closed for Specification 3.1.1.2 Moderator Temperature, Coefficient BOL and EOL limits for Specification 3.1.1.3 Boron Dilution Alarms for Specification 3. 1.2.7 Movable Control Assemblies CEA Position for Specification 3. 1.3. 1 Regulating CEA Insertion Limits for Specification 3. 1.3.6 Part Length CEA Insertion Limits for Specification 3. 1.3.7 Linear Heat Rate for Specification 3.2. 1 Azimuthal Power Tilt T for Specification 3.2.3 DNBR Margin for Specification 3.2.4 Axial Shape Index for Specification 3.2.7 Boron Concentration (Mode 6) for Specification 3.9. 1 6.9. 1. 10 The analytical methods used to determine the core operating limits shall, be those previously reviewed and approved by the NRC in:

'a ~

b.

c ~

d.

"CE Hethod. for Control, Element Assembly Ejection Analysis, "CENPD-0190-A, January 1976 (Methodology for Specification 3.1.3.6, Regulating CEA Insertion Limits).

"The ROCS and DIT Computer Codes for Nuclear Design,"

CENPD-266-P-A, April 1983 [Methodology for Specifications 3.,1.1.,1,.

Shutdown Hargin Reactor Trip Breakers Open; 3. 1. 1.2, Shutdown Margin Reactor Trip Breakers Closed;

3. 1.1.3, Moderator Temperature Coefficient BOL and EOL limits; 3. 1.3.6, Regulating CEA Insertion Limits and 3.9. 1, Boron Concentration (Mode 6)].

"Safety Evaluation Report related to the Final Design of the Standard Nuclear Steam Supply Reference Systems CESSAR System 80, Docket No.

STN 50-470, "NUREG-.0852 (November 1981),. Supplements No.

1 (March 1983),

No.

2 (September 1983),

No.

3 (December 1987)

(Methodology for Specifications 3.1. 1.2, Reactor Trip Breakers Closed; 3.1.1.3, Moderator Temperature Coefficient BOL and EOL limits; 3. 1.2.7, Boron Dilution Alarms; 3. 1.3. 1, Movable Control Assemblies CEA Position; 3. 1.3.6, Regulating CEA Insertion Limits;

3. 1.3.7, Part Length CEA Insertion Limits and 3.2.3 Azimuthal Power Ti1t Tq) e "Hodified Statistical Combination of Uncertainties,"

CEN-356(V)-P-A Revision 01-P-A, May 1988 and "System 80 Inlet Flow Distribution,"

Supplement 1-P to Enclosure 1-P to LD-82-054, February 1993 (Methodology for Specification 3.2.4, DNBR Margin and 3.2.7 Axial Shape Index).

PALO VERDE UNIT 1 6-20a Nd 66.~,6

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS i(EPORTi. (Continued) imi e.

"Calculative Methods fair 'the CE Uarge Break LOCA Evaluation Model for the Analysis of iCE and M Designed. NSSS,"-

CENPD-132, Supplement 3-P-A, June 1!985 (Methodology for Specification 3;2;1, Lini ar Heat'ate).

"Calculative Methods for the CE Smal.l. Break LOCA Evaluation Mod~1,"

CENPD-137-,P, August 1974 (MethodoTogy for'lpecification 3.2.1, Linear. flea't Ratie).

9 ~

Calculative Methods for the CE,Small Break LOiCA Evaluation Model,"

CENPD-'137-IP, Supplement

'1P, January 1977 (Methodology for Specific:ation 3,.2. 1, Linear Heat IRate)..

h.

Letter:

0.

Oi. Parr (NRC) to.F.

M. Stern (CI=)-, dated June 13, 1975 (NRC Staff Review of the Combustion Engineering ECCS Fvaluatibn Model).

NRC approval for:

6.9.1. 10f.

Letter:

K,. Kniel (NRC) to 'A'.'. Scherer (Cl:), dated September 27,'1977 (Evaluation of Topical Repio&s CENPD-133, Supplement 3-P arid CENPD-137, Supplement 1-P).

NRC approval for 6.9.1.1O.g.

J.

"Fuel Rod Maximum Allowable Preisst'>re," 'E'N-372'-P-A, May, 1990 (Methodology for Specification '3.2.1',

i';inear Heat,. Rate).

k.

Letter:

A., C. lhadani (NRC) to A., E. Scherer (CE), dated April 10, 1990, ("Acceptance fair IReferdncie CE Topical Report CEN-3/2-P",').

NfkC approval for 6.9.1..10.j.

The core operating limits shall be determ'in0d 'so'that all aipplicable limits (e;g., fuel thermal-mechah'ical limits, core them>al-hydraulic limits, ECCS limits, nuclear limits.such as shutdown mar)in, And'r'ansient" and analysis'imits) of the

. afety ainalysis are met.

The CORE OPERATING LIMITS REIPORT',

in'cludi'ng 'an'y n'kaid'-cp'rclie'evisions or supplements

thereto, shall bie provided upon issuance, for eaclh reload cycle, to the NRC Document Contrail Desk with copies to the Regional Administrator and Resident Inspector.

PAL'0 VERDE UNIT 1

6-206 Amendment, No. 69M', 101

(4pe REGS

~o Cy

'O 1Vl v'O

+w*w+

UNITED STATES NUCL'EAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 A

0 PU S

C PA DOCKET NO STN"50-529 PALO UC EA G

G STA ON UN T NO.

2 NT OF T

G S

Amendment No.

98 License No.

NPF-51 The Nuclear Regulatory Commission (the Commission).has found that:

A.

The application for amendment by the Arizona Public Service Company (APS or the licensee) on behalf of itself and the Salt River Project Agricul.tural Improvement and 'Power District, El Paso Electric Company, Southern.,Cal.i.fornia Edison

Company, Publ..ic Service Company of New Mexico, Los Angeles Department of Water and

Power, and Southern California Public Power Authority-dated February 1,

1996, 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. acti.vi.ties, authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

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

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

2.

Accordingly, the license is amended by changes

.to the Technical

,Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Operating License No. NPF-51 is hereby amended to read as follows:

iO iO

)

(2) c i

1 S ecif cati ns d

t rot etio Pl n

The Technical Specifications contained in Appendix A, as-revised through Amendment No. 98 and the Environmental-Protection=Plan contained in Appendix B, are hereby incorporated into this license.

APS shall operate the facility in accordance with the Technical Specifications and the Environmental Protection

Plan, except where otherwise stated in specific license conditions.

3.

This license amendment is effective as of the date of issuance to be implemented within 45 days of issuance.

'OR THE NUCLEAR. REGULATORY-COMMISSION har. es R. Thomas, Project Manager Project Directorate IV-2 Division of Reactor Projects. I,II/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

April 30, 1996

0 ill I

~

98 DOC S

A

.S 0-S 0.

P -5 Replace the following pages of,:the Appendix-:A.. Technical: Speci.fications"with.

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.

~NS V*

VI 2-5 B 2-4 3/4 1-1 3/4 3-23'.

3/4 3-27 3/4 5-3 3/4 5-4 3/4 5-7 8 3/4 3-2 B 3/4 5-2 B 3/4 5-3 6-20a V

VI 2-5 B 2-4 3/4 1-1 3/4 3-23 3/4: 3-27 3/4 5-3 3/4 5-4 3/4 5-7, B

3/4'-2'B/4'- :'5-2 B:3/4 5-3 6-20a

No changes were:made, to these pages,. reissued to become overleaf'ages.

i'll 1

,INDEX.

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION 3/4. 2 POWER DISTRIBUTION LIMITS

,PAGE 3/4. 2. 1 3/4. 2. 2 3/4. 2. 3 3/4. 2. 4 3/4.2. 5 3/4.2. 6 3/4.2. 7 3/4. 2. 8 LINEAR HEAT RATE.

PLANAR RADIAL PEAKING FACTORS - F...

AZIMUTHAL POWER TILT " T.............

DNBR MARGIN...........................

~

~ ~ ~

~

~ ~

~ ~ ~ ~ ~ ~ \\ ~

CS FLOW RATEo

~

~ ~

~

~ ~

~

~ ~ ~

~

o ~ o ~ ~

~ ~ ~

~

~. ~ ~ ~ ~

~ ~

~

~ ~

~ ~ ~

R REACTOR COOL'ANT COLD LEG TEMPERATURE..............--....

AXIAL SHAPE 'INDEX........................-..-.-.-..-....

PRESSURIZER PRESSURE.....

3/4 2-1 3/4 2-2 3/4 2-3 3/4 2-5 3/4 2-6 3/4 2-7 3/4 2-9 3/4 2"10 3/4. 3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION......................

3/4 3-1 HOT STANDBY.........;........

HOT SHUTDOWNS ~

~

~ ~

~ ~ ~ ~

~ ~

~

~ ~ ~

~ ~

~

COLD SHUTDOWN -

LOOPS FILLED....................

COLD SHUTDOWN " LOOPS NOT FILLED..........

~

~ '

4 ~ ~ '0 ~ '

3/4. 3. 2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION..

3/4. 3. 3 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION'.;......;...........

INCORE DETECTORS......................................

SEISMIC INSTRUMENTATION...........~......-.............

~.

METEOROLOGICAL INSTRUMENTATION.'-....

REMOTE SHUTDOWN SYSTEM............

POST-ACCIDENT MONITORING INSTRUMENTATION..- -. - - -. - - -.

'LOOSE-PART DETECTION INSTRUMENTATION.

EXPLOSIVE GAS MONITORING INSTRUMENTATION.....-. - -.. --

3/4.4 REACTOR COOLANT SYSTEM 3/4.4..1 REACTOR COOLANT, LOOPS AND COOLANT CIRCULATION, STARTUP AND. POWER OPERATION.........................,.......

3/4 3"17

'3/4'-37 3/4 3-41 3/4 3-42 3/4 3-45 3/4 3"48 3/4 3-57

. 3/4.3-61 3/4'-63 3/4 4-1 3/4 4-2 3/4 4-3 3/4 4-5 3/4 4-6 PALO VERDE - UNIT 2 V

AMENDMENT NO. 8, N~

H T G

CO D

ON~'OR~0'ELATION AND: SURVE LL N E

E UIREHENTS KCHQH 3/4.4.2 SAFETY 'VALVES HHK SHUTDOWN...........-o......,.........,...,...,..

OPERATING..... o....,.

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3/4.4'.3 PRESSURIZER PRESSURIZER...o....o................

AUXIILIARYSPRAY.....,............,...

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3/4 4-7 3/4 4-8 3/4 4-9 3/4 4-10 3/4.4.4 STEAH GENERATORS 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS.......,....

OPERATIONAL LEAKAGE.............,....

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3/4 4-11 3/4 4-18

'.3/4 4-19 3/4.4.6 3/4.4.7 CHEMISTRY o

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SPECIFIC ACTIVITY....-...................,......

3/4 4-22 3/4 4;25 3/4.4. 8 PRESSURE/TEMPERATURE LIMITS REAC"RIOR COOLANl SYSTEM........'...>>..'.

PRESSURIZER HEATUP/Cooil.OaWN LIMI1S..

OVERPRESSURE PROTIECTION SYSTEHS.l-...

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3/4 4-28 3/4 4i31 3/4 4-32 3/4.4.9 STRUCTURAL INTEGRITY................,...

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3/4.4.10-3

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REACTOR COOLANT SYSTEM VENTS.....~..i...~...'...............

CY'~0( (l(h((UhlG.,'SYS~TI(S ~ECCSQ 3/4.5.1 SAFETY INJECTION TANKS...........'.. I..'...'.'..

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3/4.5.2 3/4.5.3 ECCS SUBSYSTEMS OPERATING....;.'..

ECCS SUBSYSTEHS' SHUTDOWN.......'..>>

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3/4.5.4 REFUELING WA'llER TANK............... <....

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3/4 4-34 3/4 4-35 3/4 5-1 3/4 5-3 3/4 5-7 3/4 5-8 PALO VERDE UNIT 2 Amendn'ient No.

98

C INT M

(C ti d)

B (1)

Trip may be manually bypassed above 10 X. of, RATED THERMAL POWER; bypass shall Pe automatically removed when THERMAL POWER. is less. than or equal to 10 X'f RATED THERMAL POWER.

(2)

In 'MODES 3-4, the value may be decreased manually, to a minimum of 100

- psia, as pressurizer pressure is reduced, provided:

(a) the margin between the pressurizer pressure and this value is maintained at less than or equal to 400 psi; and (b)

.when the RCS cold leg temperature is greater than, or..equal to 485 degrees F, this value is maintained at least 140 psi'greater than the saturation pressure corresponding to the RCS cold leg temperature.

The setpoint shall be increased automatically as pressurizer pressure is increased until the trip setpoint is reached.

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

(3)

In. MODES 3-4, value may be decreased manual.ly 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)

X of the distance between steam generator upper and lower level wide range instrument nozzles.

(5)

As stored within the Core Protection Calculator (CPC).- Calculation of the trip setpoint includes measurement, calculational apd processor uncertainties.

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

PALO VERDE UNIT 2 2-5 Amendment No. 49, 98.

00 A

~I ili

,T'ABLE 2. 2-1.(Contiinued).

REACTOR.PitOTECT1VE IN!iTRUMENTATIONTRIP SETPOINT LIMITS; TAE(LE C)TATII3NS (Continued)i (6)

RATE fs the maximum rate of. diicrease iof the.!trip setpo4nt.

There are na v estHctfons on the rat'e -at wt>hach th>> setpo4,nt, can.4nLreaie.

FLOOR

.$ s the m$ n4mum-value* of the"tr4p-setpofnt.

iKNWfs the amount by i~h$ ch the 4rfp isetpofnt"fs 'belm the'- input signal

~un ess limited by Rate or,Floor.

Setpoints are based on st'an generator i4$ fferentkal pressuie.'7)

The setpo$ nt may. be altered to disable trip function during test$ ng

'pursuarit to Spiiclfkcat'fan 3;10;3..

(8), 'RATE ts the aax'fmum -rate of increase iof the tHp setpofnt.

(The rate alt

~orch the seipo$ nt can decrei<>'e'$s'o s1'~'r than five percent pir sieccInd.)

CEILING Cs the maximum value i>f the trCp setyo4nt.

~AN~is the amount by, i&fch the tHp lsittpofnt 1s above the steady 'state input signal-unless 1<mftzd.by the rata, or the calling.

(9),

X of the distance between steam genes'itor upiper and lower level narrow range Instrument Aozzles..

.-2 5

BASES REACTOR TRIP SETPOINTS (Continued)

The methodology for the calculation of the PVNGS trip setpoint values,

'plant protection system, is discussed in the CE Document No. CEN-286(V), Rev.

2, dated August 29, 1986.

Manual Reactor Tri The Manual, reactor. trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

Variable Over ower Tri A reactor trip on Variabl'e Overpower is provided to protect the reactor core during rapid, positive reactivity addition excur sions.

This trip function will trip the reactor when the indicated neutron flux power exceeds either a

rate limited setpoint at a great enough rate or reaches a preset ceiling.

The flux signal used is the average of three linear subchannel flux signals originating in each nuclear instrument safety channel.

These trip setpoints are provided in Table 2.2-1.

Lo arithmic Power Level - Hi h The Logarithmic, Power Level - High trip is provided to protect the'ntegrity 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 unless this trip is manually bypassed by the operator.

The operator may manually bypass this trip when the THERMAL POWER,level is above 10-4X of'ATED THERMAL POWER; this bypass is automatically removed when the THERMAL POWER level'ecreases to 10-4X of RATED THERMAL POWER.

Pressurizer Pressure - Hi h 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 below the nominal. lift.;setting.- of the pressurizer safety valves and its operation minimizes the undesirable opera-tion of the pressurizer 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 decrease i'

Reactor Coolant System inventory and in the event. of. an increase..in heat PALO VERDE " UNIT 2 B 2-3 AMENDMENT NO.

19

J-((

(" ')

.removal by the secondary sy. tern.

During normal ~operation, this trip's setpoint may be manually. decreased, to a minimum valiuei os 100 psia, as pressurizer pressure is reduced during plant shutdlowns, provided the margin-between

the, pressurizer pressure and thiis 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 i,s reached.

The setpoint,must also be maintained at least 140 -psi greater than th'.-sat'uration pressure: corresponding to.the RCS cold leg temperature whenever the RCS, cold leg. temperature is equal to or greater than 485 degrees F.

This will ensure safe,ty injection actuation prior to reactor vesse1I upper Ihead void formation in event of'CS

-depressurization caused by a steam 'line breaks

'These, are indicated val,ues that include allowance.. for uncertainty.

The operator may manually bypass this tr'ip when pressurizer pressure is below 400 psia.

This 'bypa'ss. is automatically removed when the pressurizer pressure increases to 500 psia; t'g, -~gh The Containment Pressure Hligh trip provides 'assurance. that a reactor trip is initiated in the event of containment building piressurization due to a pipe break inside the containment building.

The sdtpbint for this trip i0 identical to the

. afety injection setpoint.'

ow The Steam Generator Pressure Low trip provides protection i'- the event of an increase in heat removal by the secondarily: system and subsequent cooldown of the-reactor coolant.,

The setpioint is sufficiently below the full lo'ad'perating point so as not to interfere with normal operation, but still hi'gh enough to provide the required protection in the eVent of excessively high steam flow.

This trip's setpoint may be manu411;g decreased as steam generator pressure is reduced during plant shutdowns, provided the margin between the steam generator pressure ancl tlhis trip's setpoint is maintained at 'less, than, or equal to 200 psi; this setpioint increases automatically as steam generator pressure increases until the normal pressure tri,p setpoint is reached.

~R-" I-M" The Steam Generator Level Low trip provides protect'ion against a loss of feedwater flow incident and assuires that the design piressure of'he Reactor Coolant System wi'il.not be exceedled due to a decrease in heat removal by the secondary system.

This specifiedl setpoint provides allowance that there will be sufficient water inventory in the steam 'gener'ator at 'the time of the trip to provide a margin of at least 10 minutes before auxiliary feedwater is-required to prevent degraded core ciooling.

(

MIJllh The Local Power Density - High trip is provided to prevent the l.inear

'heat rate (kW/ft) in the l.imiting fuel rod in the core from exceeding the fuel design limit in tlhe event, of any design bases anticipated operationa'l occurrence.

The local power den. i.ty is calculated, in the. reactor protective, system utilizing the..'following information:

PALO VERDE UNIT 2 B 2-4 Amendment Nai.

SYSTEMS C 0 R A ERS OP N**

ITING 0

ION FOR OPERATION'.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal, to that specified in the CORE OPERATING LIMITS REPORT.

MODES 3, 4* and 5* with the reactor trip breakers open.**

KKEH:

With the SHUTDOWN MARGIN less than that specified in the CORE OPERATING LIMITS

REPORT, immediately initiate and continue boration at greater than or equal to 26 gpm to reactor coolant system of a solution containing greater than or equal to 4000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SU VEILLANC RE UIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall-be determined to be greater than or equal to that specified in the CORE OPERATING LIMITS REPORT at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of"at least the following factors:

1.

Reactor Coolant System boron concentration, 2.

CEA position, 3.

Reactor Coolant System average temperature, 4.

Fuel burnup based on gross thermal energy generation, 5.

Xenon concentration, and 6.

Samarium concentration.

4.1.1. 1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within 2 1.0X delta k/k at least'nce per 31 Effective Ful,l Power Days (EFPD).

This comparison shal.l consider at least those factors stated in Specification 4. 1. 1.;1. 1, above.

The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60. EFPD after each fuel loading.

4. 1. 1. 1.3 With the reactor trip breakers open** and any CEA(s) fully or partially withdrawn, the SHUTDOWN MARGIN shall be verified within one hour after detection of the withdrawn CEA(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the CEA(s) are withdrawn.

See Special Test Exception 3. 10.9.

- The CEA drive system not capable of CEA withdrawal.

PALO VERDE UNIT 2 3/4 1-1 Amendment No. 43-,86,98

~RA TITY CCNNIMI

~CNN RAR I RR I'IIR T E

.'I CI 'I 0

I AI 3.1.1.2 a.

The SHUTDOMN'INRGII'I shall'b'e greater "than or equal to that specified in the,CiDRE.OIPERATING LIHITS.;REPORT.;.,and, b.

For T,~ less than or equa'1 to 500'F K, sha111 be less-thain 0.99.

PP ICAB closed**.

~AC OM:

a ~

b.

.Reactor criticality shalil'ot be achieved with shutdown, group CEA movement.>>E;...I, d.

'IIA td. :

I.R:I'ith the SHUTDOWN "HARIGINI less than that specified in the CORE OPERATING LIHITS REPORT, immediately initiate and cont'jnue boration at grdeater'hYain or equal to 26'Egpm to tllie reactor coolant system of a solution contaiining greater thar> or equal to 4000 ppm,.boron'r equivalebit until the required SIIUTDOWN.HARGIN is'r'estored, And With T.d, less than.or equal-',to--500;,F and, K.i. greater than -or -equal to 0.9!),

immediatel,y vary CEA positions and/or initiate and cohtinuk boration at'reater than or equal to 26'g'pm to the. reactor coolant system of a solution containing grd.'at6r than 'or equal to 4000 ppm boron or equivalent unti't'he re~quireld Y~.> is 'restored.

SURVEILLANCE RE UIRIEHENTS 4.1.1.2.1

'Mith the reactor. trip breakets closed~+,

the, SHUTDOMN:HARGIN shall

'e determined to be greater than or equal to 'thakt 'specified in the CORIE OPERATING LIHITS REPORT:

a.

Mithin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . after detection of an inoperable CEA(s) and at least once per,12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter w'hile 'th<'. C'EA(s) is iinoperable.

~ See Special Test Exceptioins 3.10.1

'and 3',.10.9.,

~~The CEA drive system capalble of CEA withdra'wal.

PALO VERDE - UNIT 2 3/4 '1-2 Amendment No. -'.S-,55P6

B 3 3-Continued TUR S ACTUAT ON SYSTEM'STRUM NTAT 0

~TAB OTATIOAB (a)

In MODES 3-4, the value may be decreased manually, to a minimum of 100

psia, as pressurizer pressure is reduced, provided:

(i) the margin between the pressurizer pressure and this value is maintained. at less than or equal to 400 psi; and (ii) when the RCS cold'eg temperature is grea'ter, than or equal to 485 degrees F, this value is maintained: at least 140"psi greater than the saturation pressure corresponding

.to the RCS cold leg temperature.

The setpoint shall be increased automatically as pressurizer pressure is increased until the trip setpoint is reached.

Trip may be manually bypassed below 400 psia; bypass shall be automatical:ly removed whenever pressurizer pressure is greater than or equal to 500 psia.

(b)

In MODES 3-4, the 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.

(c)

Four channels

provided, arranged in a selective two-out-of-,four configuration (i.e., one-out-'of-two taken twice).

'(d)

The proper two-out-of-four combination.

(e)

Input to channels.

The provisions of Specification 3.0.4 are not applicable.

C ON S

T M NTS ACTION 12 ACTION 13 Steam Generator Pressure Low Steam Generator Level I-Low (ESF)

Steam Generator Level. 2-Low (ESF)

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 .

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

If the inoperable channel is

bypassed, the desirability of,maintaining thi's channel in the bypassed condition shall be reviewed in accordance with Specification 6.5..1.6.g.

The channel-shal.l be returned to OPERABLE,status,no'ater than during the next COLD SHUTDOWN'.

With a channel process measurement circuit that affects multiple functional uni'ts inoperable or in test, bypass or trip all associated functional units as listed below.

Process Measurement Circuit 1.

Steam Generator Pressure-Low 2.

Steam Generator Level (Wide Range)

Steam Generator Level Low (RPS)

Steam Generator Level I-Low (ESF)

Steam Generator Level 2-Low (ESF)

PALO VERDE UNIT 2 3/4 3-23 Amendment No. 84,98

TABLE 3.3-'.3 (Cnntinn!!d)'NGINEERED, SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION ACTION STATEMENITS ACTION 14-ACTION 15-ACTION 16-ACTION 17-ACTION 18-With the number of channels OPERABLE one less than the Ninimum Channels

OPERABLE, STARTUP and/or POWER OPERATION.may co'ntinuie provided the foiled>>ing 'conditions are

. atisfied:

a.

Verify tlhat one of the inoperable channels has been bypassed and place the othe>>

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.

1.

Steam Generator Pressure-Low Functional Unit Bypassed/T>ripped Steam Generator Pressure - Low Steajm Generator,Level 1 - Low (ESF)

Steam Generator Level 2 - Low (ESF)

WitlI>> the numbe>> of OPERABLE channels oI>>e less"than the Minimum Numlber of Channels, operation may cont'inue for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

After. 6 hou>>s ope>> ation may continue provided at least 1 t>>ain of essential filtration is in operation, otherwise, be in HOT STANDEIY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COULD SHIUTDOWN within the, following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

2.

Steam.Generator Level - Low Steam Generator Level - Low (RPS)

(Wide Range)

Steam Generator Level 1 - Low (f:SF)

Steam Generator Level 2 - Low (ESF)

STARTUP and/or.

POWER OPERATION i'nay'bntinde until. the pe>>'formance of the next required CHANNEL FUNCTIONAL'EST.

Subsequent,.

STARTUP and/or POWER OPERATION may. cbntinise'-if one channel is restored to OPERABLE status and the provisions of ACTION 13 a>>'e satisfied.

With. the number of OPERABLIE channels one 'less than the Total Number o1F Channels restore the inoperable channel to OPERABLi-:

status w'ithin 48 hIIurs 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 'ith the number of OPEIVjBLE channels one 'less than the Total Number o1F Channels be i'n 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 H01I SHUTDOWN within tlie following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ;

however, one channel may be bypassed for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for surveillance testing provided the other chahnel is OPERABLE.

With the number of OPERABLE channels one less than the Minimum Number of Channels,,

restore thei inoperable channel to OPIERABLIE sta'tus within 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months s-.or 'be in at least,.lH01'.

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 .

PALO VERDE - UNIT 2 3/4 3~24',

'-'"(" I((3833-3 (3'l 3(""

ATUR S C

U TIO SYST M

NST U

A ON

~TAB(

3 TAT Bll (I)

In MODES 3-4, the value may"be decreased..manual:ly,,

to a minimum of 100

psia, as pressurizer pressure is, reduced, provided:

(a) the margin between the pressurizer pressure and this value is maintained at less than or equal to 400 psi; and (b) when the RCS cold leg temperature is greater than or equal to 485 degrees F, this value is maintained at least 140 psi greater than the saturation pressure corresponding to the RCS cold'eg temperature.

The setpoint shall be increased automatically as pressurizer pressure is increased until the trip setpoint is reached.

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

(2)

X of the distance between steam generator upper and lower level narrow range instrument nozzles.

(3)

In MODES 3-4, 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 automati'cally as steam generator pressure is increased until the trip setpoint is reached.

(4)

X of the distance between steam generator upper and lower level wide range instrument nozzles.

PALO VERDE UNIT 2 3/4 3-27 Amendment No. 98

~I 41

G' NG IO PE RATION 3.5.2 Two independent Emergency Core -Cooling-System-=(ECCS) subsystems shall be OPERABLE with each subsystem comprised of:

a.

One OPERABLE high-pressure safety injection pump, b.

One OPERABLE low-pressure safety injection pump, and c.

An independent OPERABLE flow path capable of taking suction from the refueling water tank on a safety injection actuation signal and automatically transferring suction to the containment sump on a recirculation actuation signal.

MODES I, 2, and 3*.

~AC ~0:

a ~

Mith one ECCS subsystem inoperable, restore the'noperable subsystem to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at-;least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT...SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

In the event the ECCS is. actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted,to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

The current value of the usage factor for each affected injection nozzle shal:1'e--provided in this Special Report whenever its value exceeds 0.70.

Mith pressurizer pressure greater than or equal to 1837 psia, or RCS cold leg temperature greater than or equal to 485 degrees F.

PALO VERDE UNIT 2 3/4 5-3 Amendment No. 58,98

K <<

~E<<E 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a.

At least once per 12,hours by verifying that the Following valves are in the indicated."positions with thie val've.'>'key-locked. shut:.

Valve Number Vallve-Fun'ction Valve Pos.ition 1.

SIA HV-6O4 1.

HO'T L,EG INJECTION I.

SHUT 2.

SIC HV-321 2.

HOT LEG INJECTION 2.

SHUT 3.

SIB HV-6O9 3.

HOT LEG.INJECTION 3.

SHUT 4.

SID HV-331 4.

HO I LEG INJECTION 4.

SHUT b.

At least once per 31 days by:

1.

Verify'ing that each valve (manual, power-operated, or automatic) in the flow path that is not '~locked,

'sealed, or otherwise secured in position, is in its c'orrect position,

.and 2.

Verifying that the ECCS piping is full of.water.by'venting the accessible discharge piping

-high'6ints.'.

By a visual inspection which verifies'hat no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the cor>tainment sump and cause restriction of th'e pump suctions, during LOCA condit,ions.

This visisal inspection shall be performed:

1.

For all accessible areas of the'ontainment 'prior'to establishing CONlAIINMENT INTEGRITY, and 2.

At lea.st once daily of the affected areas within containmeint by containment entry and during the final entry when CONTAINMENT INTEGRITY i. established.,

d.

At least once per 18 months by:

PALO VERDE UNIT 2 3/4 5-4 AMENDMENT NO 66

M G

C R

00 NG YST MS UBSYST M

SHUTDOWN M T CO T ON FOR OP RATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a.

An OPERABL'E high pressure safety injection pump,.and b;

An OPERABLE flow path capable of taking suction from the refueling water tank on a safety injection actuation signal and automatically transferring suction to the containment sump on a recirculation actuation signal.

MODES 3* AND 4.

a ~

With no ECCS subsystem OPERABLE, restore at least one ECCS subsystem to OPERABLE status within I hour or be.in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b.

In the event the.ECCS is actuated and injects water, into, the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to-Specification 6.9.2 within-90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

SURVEI LANCE RE UIREMENTS 4.5.3 The ECCS subsystem shall be demonstrated OPERABLE per the applicable surveillance requirements of Specification 4.5.2.

With pressurizer pressure less than 1837 psia and RCS cold leg temperature less than 485 degrees F.

PALO VERDE UNIT 2 3/4 5-7 Amendment No. 98

igi

~,

~

ASES d

4.3.

A TOR RO T

ST N

U T

0 GIN The OPERABILITY of the reactor protective and'ngineered.

Safety Features Actuation Systems instrumentation and bypasses"ensures that.(I). the. associated.

Engineered Safety Features Actuation action and/or reactor trip will be initiated

'hen 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 safety analyses.

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-A, "RPS/ESFAS Extended Test Interval Evaluation,,"

and CEN-327-A, Supplement I, and calculation 13-JC-SB-200-Rev.

01.

Response

time testing of resistance temperature

devices, which are a part of the reactor protective system, shall be performed by using in-situ loop current test techniques or another NRC approved method.

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'actors and"CEA 'devi'ation 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 design of the Control Element Assembly Calculators (CEAC) provides reactor protection in the event one or both CEACs become inoperable.

If one CEAC is in test or inoperable, verification of CEA 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 CPCs in conjunction with: plant Technical Specifications will use DNBR and LPD"penalty factors-and 'increased DNBR and LPD margin to restrict reactor operation to a power level that will ensure safe operation of the plant. If the margins are not maintained, a reactor trip will occur.

The value of the DNBR in Specification 2.1 is conservatively compensated for measurement uncertainties.

Therefore, the actual RCS total flow rate determined by the reactor coolant pump differential pressure instrumentation.,or by calorimetric calculations does not have to be conservatively compensated for measurement uncertainties.

PALO VERDE - UNIT 2 B 3/4 3-1 AHENDNENT NO. 49, 64

S S

AC

~ML~Q~N~~ SA~FT~FFQU~RS ACTUATION SVSI'EM (Continued)

The measurement of'esponse time at, thee. speci, fied. frequencies'rov'ides

,'ssurance that the protective and E.iF actioia f'unction associated with each channel is completed within the time limit assumed in the safety analyses.

No credit was taken in the anallyses for thole ich6nn'els with response times indicated as not app'Iicable.

1'he instrumentatioh r'espon'se'times are made. up of the time to generate the trip signal at the d&te'ctor. (sensor response time) and the time for the signal to interrupt poWer'6 the'CEA drive mechanism (signal or trip delay time).

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) uti'lizing repllacems.nt sensors with certified response times.

During normal operation,,

the low pressurizer pressure trip setpoint may be manually decreased, to a roinimum value of'00 psia, as pressurizer pressure is reduced during plant shutdowns,,

provided the miargin between the"pressurizer pressure and this trip's setpoint is maintai'net'i a't les's than or equal to 400 psi; this setpoint increases automatically as pressi>rizer pressure increases until the trip setpoint is reached.

This se'tp6int must also be maintained at least 140 psi greater than the saturation pIes4ure dor0e~ponding ta the RCS cold leg temperature whenever the RCS cold leg te'mpeature is equal to or greater than 485 degrees F.

This wi'll ensur'e Safety ihjdction actuation prior to reactor vessel upper head void formation lin.levtent -of RCS;. depressurization caused by a steam'ine break.

These are indicated values that include allowances for uncertainty.

The operator maiy titian'ually"bypass 'the low

'pressurizer pressure trip ~when pressurizer p'ressure is below 400 psia.

This bypass is automatically removed when the presstirizer pressure increases to 500 psia.

PALO VERDE - UNIT 2 8 3/4 3-2 Amendment,No.

VJ-,75, 98

3/4.5 EMERGENCY CORE COOL'ING SYSTEMS (ECCS)

BASES 3/4. 5. I SAFETY INJECTION TANKS The OPERABILITY of each of the Safety Injection. System"(SIS) -safety injection tanks ensures that a sufficient volume of borated. water:will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure fal.ls below the pressure of the safety injection tanks.

This initial surge of water into the RCS provides the initial cooling mechanism during large RCS pipe ruptures.

The limits on safety injection tank volume, boron concentration, and pressure ensure that the safety injection tanks will adequately perform their function.in the event of a LOCA in MODE 1, 2, 3, o'r 4.

A minimum of,25K narrow range corresponding to 1790 cubic feet and a

maximum of 75K narrow range corresponding to 1927 cubic feet of borated water are used in the safety analysis as the volume in the SITs.

To allow for instrument accuracy, 28X narrow range corresponding to 1802 cubic feet and 72K narrow range corresponding to 1914 cubic feet, are specified in the Technical Specification.

A minimum of 593 psig and a maximum pressure of 632 psig are used in the safety analysis.

To allow for instrument accuracy 600 psig minimum and 625 psig maximum are speci, fied in the Technical Specification.

A boron concentration of 2000 ppm minimum and 4400 ppm maximum are used in the safety analysis.

The Technical Specification lower limit of 2300 ppm in the SIT assures that the backleakage from RCS will not dilute the SITs below the 2000 ppm limit assumed in the safety analysis prior to the time when draining of the SIT is necessary The SIT isolation valves are not single failure pro'of; therefore, whenever the valves are open power shall be removed from these valves and the switch keyl'ocked open.

These precautions ensure that the SITs are avai.lable during a Limiting Fault.

The SIT nitrogen vent valves are not single failure proof against depressurizing the SITs by spurious opening.

Therefore, power to the valves is removed while they are closed to ensure the safety analysis assumption of four pressurized SITs.

All of the SIT nitrogen vent valves are required to.be operable so that, given a single failure, all four SITs may still be vented during post-LOCA long-term cooling.

Venting the SITs provides for SIT depressurization capability which ensures the timely establishment of shutdown cooling entry conditions as assumed by the safety analysis for small break LOCAs.

The limits for operation with a safety injection tank inoperable for any reason except an,isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional safety injection tank which may result in unacceptable peak cladding tempera-tures.

If a closed isolation valve cannot be immediately opened, the: full capability of one safety injection tank is not available and prompt action is required to place the reactor in a MODE where this capability is not required.

For MODES 3 and 4 operation with pressurizer pressur e less than 1837 psia PALO VERDE UNIT 2 B 3/4 5-1

~!I(( M(III((M&('(2 ASES

(((((((.((

the Technical Speci f'icition. require.,a minimum of..fi7X wide range correspokding to. 1361 cubic, feet and a ma>(imum of 75iX narroiv range corresponding. to 1'92V cubic feet: of borated water per tanlk,. when three safety injection tanks are

~

inoperable and a minimum-of 36X wide !range corresponding to.908 cubic:feet and a

maximum of 75X narrow-range"co'rre,spiohding-to-1927 cubic feet. per tank; iwhkn i

four.:safety injection tanks're-.aipeIrabl'e.at (a"mi'nia'ium:pressure'f 235 psig and a maximum pressure of 625 psig.

To'l'low:for instrument';. inaccuracy,-..60X..>tide, range instrument corresponding to 1415 cubic fee't; 'and 72X harrow range instrument corresponding:to 1914 cubic feet, whein three safety injection tanks are operable, and 39X widie 'range instrument corresponifing to 962 cubic $e6t, and 72X narrow range inst!rue'ienIt corresponding to 1914 cubic 'feet, when four SITs are operable,,

are specified 1n the Teclhnical S'pecif'ications.

To allow for instrument inaccuracy"254 psig is. specified in. the-Technical.Specificatioiis.

The instrumentatiain vs.

volume correlatio>>n for'he 'SIls is as follows:

~o

~~r<i~w~el,,,

Midg" R~ne 962 ft.'OX 39X 1415 ft;,

<OX

.C)OX 1802 ft, 28X

,78X 1914 ft 72X 83X 3 Sl((.

(((((((M(((L(

The OPERABILITY of two separate and-incf'epenclent.

ECCS subsystems with the.

indicated.

RCS pressure greater than or-equal to 1837'sia, or. w'ith-.the

'indicated RCS cold"leg 'temperat;ure gIreater than or.equal tai 485.degrees F

ensures that sufficient.emergency cadre cool1ng'apability will -be available in the, event of a LOCA assumiing the 'loss of ond subsystem through any single failur'e consideration.

These indicated values iricTude.a'1 1aiwances f'r.

uncertainties.,Either subsystem ioperating 1n cor'adjunct;ion with the safety

'njection.

tanks's capable of suplplying sufficieht core,coailing..to,.limit..the peak cladding -temperatures withiin acceptabl

d. 1'imits'ear ill postul ated:break siies. ranging-from, the double-ended break of the largest-RCS:cold 'leg pipei downward.

In addition, each ECCS subsystem provides long-term core coolin'g capability in the recircul'ation miode during 'th'e aIccident recovery,. period.

'he Mode 3 safety analysis c!redlits one HPSI puInp to provide negative reactivi.ty insertion to protect; the core and RCS fol,lowing 'a steam l.i'ne'r'eak when RCS cold leg temperature is 485 degrees, F. or, greater.

Requiring two operable ECCS'ubsystems in the situatio'n 'will'6su're~'one 'HPSI pump.is available assumin'g sIlngle failure of the.other HPSI primp.

Mith the RCS cold leg temperature below 485'degrees

.F, one:OPERABLE'CCS subsystem is acceptable wiithout sing!le'ailure c6'nsideration oh the bas:i's of

'the stable reactivity condition of t,he reactor and the, l,imited core cool-ing requirements.'.

The trisodium phosphate dodecahydrate.(TSP)~

stored in dissol.ving basket.

located. in the, containment baseme!nt's,-provided'o'.miriimize...the-possibi'lity of.

corros1on cracking. of certa1n IIietal,companents during operation of thie I.:CCS'ollowing a

LOCA.

The,.TSP proi~idied this:protection by dissolving in the sump water and causing it. final. pH, to be. raised

-to greater tlhan, or equal to 7.0.

PALO VERDE - UNIT 2 B 3/4, 5-2 Amendment'o.

98

}

<<~'sPt'0 ST S

BAS.S CC S

(Continue'd)

The, surveillance requirements provided to ensure, OPERABILITY of each component ensure that at a minimum, the, assumptions used: in:.the safety analyses are met and that subsystem OPERABILITY is maintained.

Surveillance require-ments for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA.*

Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to:

(I) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection. points in accordance with the assumptions used in the ECCS-,LOCA analyses, and (3) provide an acceptable level, of total ECCS flow -to al~l injection points equal to or above that assumed in the ECCS-LOCA analyses.

In specification 4.5.2.h, the specified flows include instrumentation uncertainties.

The requirement to dissol.ve a representative sample of TSP in a sample of RWT water provides assurance that the stored TSP will dissolve in borated water at the postulated post-LOCA temperatures.

The term "minimum bypass recirculation flow," as used in. Specification 4.5.2e.3.

and 4.5.2f., refers to that flow directed back to the RWT from the ECCS pumps for pump protection.

Testing of the ECCS pumps under the condition of minimum. bypass recirculation flow in Specification 4.5.2f. verifies that the performance of the ECCS pumps supports the safety analysis minimum RCS pressure assumption at zero delivery to the RCS.

The OPERABILITY of the refueling water-tank (RWT)- as 'part--of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA.

The limits..on,RWT minimum,volume, and boron concentration ensure that (I) sufficient water plus lOX margin is avail-able to permit 20 minutes of engineered safety features pump operation, and (2) the reactor will remain subcritical in the cold condition fol.lowing mixing of the RWT and the RCS water volumes with all control rods inserted except for the most reactive control assembly.

These assumptions are consistent with the LOCA analyses.

The following test conditions, which apply during flow balance tests, ensure that the ECCS subsystems are adequately tested.

1.

2.

3.

The pressurizer pressure is at atmospheric pressure.

The miniflow'bypass recirculation lines are aligned for injection.

For LPSI system, -(add/subtract) 6.4 gpm (to/from) the 4800 gpm requirement for every foot. by"which the difference of 'RWT water level above the RWT RAS setpoint level (exceeds/is less than) the difference of RCS water level above the cold leg centerline.

PALO VERDE UNIT 2

,B 3/4 5-3 Amendment No. 29,98

EMERGENCY CORE COOLING SYSTEMS BASES REFUELING MATER TANI( (Continued)

The contained water volue liait inc14dks ln allow'ance for water not usable because of tank diischarge line location or other physical characteristics.'he liaits on contaiined water vol>me <<nd boron concentration oW the M also insure a pH vallue of between: 7.0 and 8.S fOr 't% s'oluti'on recircu'lat'ed within containoent after a LOCA.

This pH band sinieizes the evolution~of~

iodine and ainiaizes the effect of chloride and.caustic stress corrosion on aechanical systeas and cceponents.

The limit. on the RMT solution temperature ensures that 'the assuoptions used in the LOCA analyses:,

remain valid.

PAln VEROE - UNIT 2 8 3/4 5-4

ADM N ST T

E CONTROLS T

P 6.9.1.9 Co}e operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

C.

d.

e.f.

g h.

1

~

k.l.

Shutdown Hargin Reactor Trip, Breakers Open for Specification 3.1.1.1 Shutdown Hargin - Reactor Trip Breakers Closed for Specification 3.1.1.2 Moderator Temperature Coefficient BOL and EOL limits for Specification 3.1.1.3 Boron Dilution Alarms for Specification 3. 1.2.7 Movable Control Assemblies CEA Position for Specificati'on 3. 1.3. 1 Regulating CEA Insertion Limits for Specification 3. 1.3.6 Part Length CEA Insertion Limits for Specification 3. 1.3.7 Linear Heat Rate for Specification 3.2.1 Azimuthal Power Tilt - T for Specification 3.2.3 DNBR Margin for Specification 3.2.4 Axial Shape Index for Specification 3.2.7 Boron Concentration (Mode 6) for Specification 3.9. 1 6.9. 1. 10 The analytical methods used to determine the-core operating limits shall be those previously reviewed and approved::by the NRC in:

a ~

b.

co d.

"CE Method for Control Element Assembly Ejection Analysis,"

CENPD-0190-A,.January 1976 (Methodology for Specification 3. 1.3.6, Regulating CEA Insertion Limits).

"The ROCS and DIT Computer Codes for Nuclear Design,"

CENPD-266-P-A, April 1983

[Methodology for Specifications

3. l. 1.1',

Shutdown Hargin Reactor Trip Breakers Open; 3;- 1.:1.2, -Shutdown Hargin-Reactor Trip Breakers Closed;

3. 1. 1.3, Moderator Temperature Coefficient, BOL and EOL limits;. 3.1.3.6, Regulating CEA Insertion Limits and 3.9. 1, Boron Concentration (Mode 6)].

"Safety Evaluation Report related to the Final Design of the Standard

Nuclear, Steam Supply Reference Systems CESSAR System 80, Docket No.

STN 50-470, "NUREG-0852 (November 1981),

Supplements No.

1 (March 1983),

No.

2 (September 1983),

No.. 3 (December 1987).

(Methodology for Speci.fications 3.1. 1.2, Shutdown Margin. Reactor Trip Breakers Closed;

3. 1.1.3, Moderator. Temperature. Coefficient BOL and EOL limits; 3. 1.2.7, Boron"Dilution Alarms;- 3. 1.3.1,. Movable Control Assemblies CEA Position; 3.1.3.6, Regulating CEA Insertion Limits; 3. 1.3.7, Part Length CEA Insertion Limits and 3.2.3 Azimuthal Power Ti-lt Tq).

"Hodified Statistical Combination of Uncertainties,"

CEN-356(V)-P-A Revision 01-P-A, May 1988 and "System 80 'nlet Flow Distribution,"

Supplement 1-P to Enclosure. 1-P to LD-82-054, February 1993 (Methodology for Specification 3.2..4, DNBR Margin and 3.2.7 Axial Shape Index)'.

PALO VERDE UNIT 2 6-20a Amendment No. 55-,62-,46, 98

~I DMINISTRATIVE CONTROLS

))t)JTI )KB))) ) )

)

d) e.

"Calculaitive Hethods for the CE Large l)reak LOCA Evaluation Model for the Analysis of CE and M Designed NSSS,"

CENPD-132, Supplbmhnt~

3-P-A, June 1985 (Methodology for Specification 3.2.1, Linear Heat Rate).

f.

"Calculative Methods for the CE: Sr<all lfreak LOCA Evaluation Model,"

CENPD-137-.P, August 1974 (Methodology for Specification 3.2I.I)

Linear Heat Rate).

g.

h.

"Calculative Methods four the CE Srisall EIreak LOCA Evaluationl Mbddl,~

CENPD-137-P, Supplement 1P, January 1977 (Methodology for Specification 3.2.;1, Linear Heat Rat'e).

Letter:

O.

D. Parr (NRC) to F.'.'. S'tern (CE'), dated. June 13, 1975 (NRC Staff Review of the 'Combustidn lEnglinher'ing ECCS:Evaluation

'odel).

NRC approval for:

6.9.1.10f.

i.

Letter:

K. Kniel (NRC) to A.

E'. Schhrer (CE), dated September 27, 1977 (Evaluation o)F Topical Reports CE(PD-133, Supplement 3-,P and CENPD-137, Supplement 1-P).

NRd kppkoval for 6.9. 1. 10.g.

"Fuel Rod Maximum Allowable Pressure,,',"

CEN-372-P-,A, May 1990'Hethodo'logy

)For Specification 3.2. 1 Linear Heat Rate)..

Letter".

A.

C,. Thadani (NRC) to A.

E,, Scherer (CE), dated Apri.l 10, 1990, ("Acceptance for I'Ieference CE Topical Report CEN-372-P").

NRC approval for 6.9. 1.,10.j.

The core operating limits shal,l be determine'd 4o that all applicable 'limits (e.g., fuel therma',I-mechanical lioiits, core-thermal-hydraiul:ic. limits-,

ECCS limits, nuclear limits. uch as shutdown margin and transient and ana'lysis limits) of. the safety analysis, are met.

The CORE OPERATING LIMI1'S REfORT, includihg an) mid-cycle revisions or supplements

thereto, shall be provided upon 'is<ua'ncaa, for each reload cycle, to the NRC Document Control Deslk with copies'O the Regional Administrator and Resident Inspector.,

'PALO VERDE - UNIT '2 6-20b Amendment No. 5S-,'-,7489

~S RE0y (4

~o A,

ClO I

V)

O~

/p qO

++*++

UNITED STATES NUCLEAR REGULATORY COMMISSION'ASHINGTON, D.C. 20555-0001' PUBL C S

COMPA T A.

S 50-530 ND T TO F

C NG STAT 0 UN

NO.

3 0

RAT NG L C NSE Amendment No. 78 License No. NPF-74 1.

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

A.

The application for amendment by the Arizona Public Service Company (APS or the licensee) on behalf of itself and the Salt River Project Agricultural Improvement and Power District, El Paso Electric Company, Southern Cali.fornia Edison Company,,

Public Service Company of New Mexico, Los Angeles Department of Water and

Power, and Southern California Public Power Authority"dated February 1,

1996, complies with the standards arid 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 and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

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

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

2.

Accordingly, the license is amended by,changes to the Technical Specifications as indicated in the attachment to this 1-icense amendment, and paragraph 2.C(2) of Facility Operating -License No. NPF-74 is. hereby amended to read as follows:

J

~

(2) e t

ns 'on t

c The Technical Specifications contained in: Appendix A, as revised through Amendment No. 78,. and the Environmental..Protection.

Plan contained in Appendix B, are hereby incorporated into this license.

APS shall operate the facil.ity. in accordance with the Technical Specifications and the Environmental Protection

Plan, except where otherwise stated in specific license conditions.

3.

This license amendment

.i's effective as of its date of issuance to be implemented within 45 days of, issuance.

FOR THE NUCLEAR REGULATORY-COMMISSION Charles R. Thomas, Project Manager Project Directorate IV-2 Division. of Reactor.

Projects,.I-I-I/IV Offi'ce of..Nuclear, Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

APri'1 '30, 1996

~I 4l

FAC Y

CK 50-530 NS

0. NPF-74 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.

The revised pages are identified by amendment number and contain earginal lines indicating the areas of change.

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

EEHMO V*

VI 2-5 8 2-4 3/4 1-1 3/4 3-23 3/4 3-27 3/4 5-3 3/4 5-4>>

3/4 5-7 8 3/4 3-2 8 3/4 3-3 8 3/4 3-4 8 3/4 5-2 6-20a V

VI 2-5 8 2-4

'/4 1-1 3/4 3-23 3/4 3-27 3/4 5-3 3/4 5-4 3/4 5-7 8 3/4 3-2 8 3/4 3-3 8 3/4 3-4 8 3/4 3-5 8 3/4 5-2 6-20a

No changes were made to these

pages, reissued to become overleaf pages.

4I'

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION 3/4. 2 POWER'ISTRIBUTION. LIMITS PAGE 3/4.2.1 3/4. 2'. 2 3/4.2.3 3/4. 2. 4 3/4.2. 5 3/4. 2. 6 3/4.2. 7 3/4. 2. 8 LINEAR HEAT RATE..-...........

~ ~

~

~.

~

PLANAR RADIAL PEAKING FACTORS - F....,........,.........

AZIMUTHAL POWER TILT - T.....................-

~

.DNBR MARGIN..............-.-.......--...

CS FLOW RATEo ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~

~ ~ ~ ~ ~ ~ ~

~ ~ o ~ ~ ~ o ~ ~ ~

~

R REACTOR COOLANT COLO LEG TEMPERATURE.....................

AXIAL SHAPE INDEX.

PRESSURIZER PRESSURE................................

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

'3/4 2-5 3/4 2-6 3/4 2-7 3/4 2-9 3/4 2"10 3/4.3 INSTRUMENTATION 3/4. 3. 1 REACTOR PROTECTIVE INSTRUMENTATION............

3/4. 3. 2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.............................

3/4. 3. 3 MONITORING INSTRUMENTATION

~ ~

~

~ ~

3/4 3 1 3/4 3-17 RADIATION MONITORING INSTRUMENTATION.............

3/4 3-37 INCORE DETECTORS...................

3/4 3-41 SEISMIC INSTRUMENTATION..............................

3/4 3-42 METEOROLOGICAL INSTRUMENTATION.......................

3/4 3-45 REMOTE SHUTDOWN'YSTEM...............................

3/4 3-48 POST-ACCIDENT MONITORING INSTRUMENTATION...

LOOSE-PART DETECTION INSTRUMENTATION..........

EXPLOSIVE GAS MONITORING INSTRUMENTATION.............

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION........................,....,..

OT STANDBYo ~ ~

~

~ ~

~

~ ~

~ ~ ~

, ~ ~

, ~. ~ ~ ~ ~ ~ ~

. ~ ~ ~ ~

, ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~

, ~. ~ ~. ~.

H OT SHUTDOWNo ~

~

~ ~ ~ ~ ~ ~

~ ~ ~

~

~ ~ ~ ~

~ ~ ~

~

~ ~ ~

. ~ ~ ~ ~ ~

~

~ ~

~ o ~

H COLD SHUTDOWN " LOOPS FILLED;........

COLD SHUTDOWN " LOOPS NOT FILLED....

'/4 3-57 3/4 3-61 3/4 3-63

,3/4 4-1 3/4 4"2 3/4 4-3 3/4 4-5 3/4 4-6 PALO VERDE " UNIT,3 AMENDMENT NO. 88

~SQJQg 3/4.4.2 SAFETY VAI YES

SHUTDOWN,

~

~ ~ ~

~ ~

~ ~ ~ ~ ~ ~ o

~

~ ~

~ l ~ '

~

o i o ~ ~

~

<<l ~

OPERATING...,.................

<. J.. >.

3/4.4.3 PRESSURIZER

~ e

~ ~

~ ~ ~ ~

~

e PAGE 3/4 4-7 3/4 4-8 PRESSURIZER........

~.....,,......,......

AUXILIARYSPRAY.......;.............

~ ~ ~

~

~ ~ ~

~

~ ~

e 3/4 4-9 3/4 4-1O 3/4'.4.4 3/4.4.5 STEAM GENERATORS.........,...............................

REACTOR COOLANT SYSTEM LEAKAGE 3/4 4-11 3/4.4.6 3/4.4.7 LEAKAGE DETECTIONI SYSTEMS...;.......

OPERATIONAL LEAKAGE,......,..;. ~......

CHEMISTRY.............,......,...........

SPECIFIC ACTIVITY..........,...:........

~

4

~

4

~

II ~

~

~ ~ 0 3/4 4'-18 3/4 4-19 3/4 4-22 3/4 4-25 3/4.4.10 REACTOR COOLANT SYSTEM VENTS...........

~YCOU-Wl'u!'

"W>KIK 81

'/4.4.8'RESSURE/TEMPERATIJRE LIMITS REACTOR COOLANT SYS'll'EM..........'. '...'.

PRESSURIZER HEATUP/COOLDOWN LIMITS..

OVERIPRESSIJRE PROTEC'll'ION SYSTEMS.....

3/4. 4;9 STRUCTURAL INTEGRITY..

~

~ ~

~

~ ~

~

11

~

~ ~

~

~ ~, ~

~

~ ~

~

~ ~

3/4 4-28 3/4 4-31 3/4 4-32',

3/4 4-34 3/4 4-'35 3/4.5.1 3/4.5.2 3/4.5.3 3/4.5.4 SAFETY INJECTION IANKS,......,...........

ECCS SUIBSYSTIEMS - OPERATING........................,......

ECCS SUIBSYSTIEMS - SHUTDOWN........,..............

REFUELIING WATER TANK..........

3/4 5.-1 3/4 5-3 3/4 5-7 3/4 5-8 PALO VERDE UNIT 3 Amendment hlo. 78

I ~

l IAAF (c I

d) f,h REACT R

ROT TIV NSTRUH NTAT ON TR P

S TPO NT LIHITS T

NOTATIONS Trip may be manually bypassed above 10 X of RATED THERHAL POWER; bypass shall be automatically removed when THERHAL POWER is less than or equal to 10 X of. RATED THERHAL POWER.

(2)

(3)

(4)

(5)

(7)

(8)

In HODES 3-4, the value may be decreased manually,-.to-:a,minimum of 100

psia, as pressurizer pressure is reduced, provided:

(a) the margin between the pressurizer pressure and this value is maintained at'ess than or equal to 400 psi; and (b) when the RCS cold leg temperature is greater than or equal to 485 degr ees F, this value is maintained at least 140 psi greater than the saturation pressure corresponding to the RCS cold leg temperature.

The setpoint shall be increased automatically-as pressurizer pressure is increased until the trip setpoint is reached.

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

In HODES 3-4, 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 i's increased until the trip setpoint is reached.

X of the distance between steam generator upper.

and lower level wide range instrument nozzles.

As stored within the Core Protection Calculator (CPC).

Calculation of the trip setpoint includes measurement, calculatiopal and processor uncer-tainties.

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

QQg is the maximum rate of decrease of the trip setpoint.

There are no restrictions on the rate at which the setpoint can increase.

Mgg is the minimum value of the trip setpoint.

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

Setpoints are based on steam generator differential. pressure.

The setpoint may be altered to disable trip function during testing pursuant to Specification,3. 10.3.

~.is the maximum rate of increase of the trip setpoint.

(The rate at which the setpoint can decrease is no slower than five percent per second.)

~llJgq is the maximum value of the trip setpoint.

ggg is the amount by which the trip setpoint is.above the steady state input signal unless limited by the rate or the ceil.ing.

X of the distance between

.steam generator upper'and lower level narrow range instrument nozzles.

PALO VERDE UNIT 3 2-5 Amendment No. 48,78

41

'0 BASES REACTOR TRIP SETPOINTS (Continued)

The methodology for the calculation of the PVNGS trip setpoint values, plant protection

system, is discussed in the CE Document No.

CEN-286(V),

Rev.

2, dated.August 29, 1986.

Manual Reactor Tri The Manual reactor trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor'rip capability.

Variable Over ower Tri A reactor trip on Variable Overpower is provided to protect the reactor core during rapid positive reactivity addition excursions.

This trip function will trip the reactor when the indicated neutron flux power exceeds, either a

rate limited setpoint at a great enough rate or reaches a preset ceiling.

The flux signal used is the average of three linear subchannel flux signals originating in each nuclear instrument safety channel.

These trip setpoints are provided in Table, 2.2-1.

Lo arithmic Power Level - Hi h 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 unless this trip is manually bypassed by the operator.

The operator may manually bypass this trip when the THERMAL POWER level, is above 10-~X of RATED THERMAL POWER; this bypass is automatically removed when the THERMAL POWER level decreases'o lO-~X, of RATED THERMAL POWER.

Pressurizer Pressure - Hi h 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 i'. below the: nominal lift"setting-'of the

"'ressurizer sNF4ty valves and its operation mini'mizes the undesirable, opera-tion of the pressurizer safety val.ves.

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 decrease in Reactor Coolant System inventory and in, the event of an increase in: heat PALO VERDE " UNIT 3 B 2-3 AMENDMENT NO'8

BASES Pre s ri er, P.

e sure -~aiw (Continued) removal by the seconda'ry sy:stem.

During normal operation, this trip's setpoint.may be manual'ly, deicreased,.

to a minimum valuie of 100 psia, as pressurizer pressure is reduced duririg plant shutdownis, prov;ided the margin between-the pressurize'r piressure and this triIp's setpoint is maintaineql aIt less than.or equal to-400"psi; this setpoirit -increases"automaticallly',"as pressurizer

pressure, iihcreases until the trip -setpoint. 'is reached; The

.setpoint must..also. be. maintained at:least..140.psi;.ijreater'than.-.,the.,saturation

,pressure correspoinding tai the RCS cold leg temp'erasure whenever the RCS cold leg temperature is equal to or greater, tlhan 485 degrees F.

II'his will dnsiir6 safety injection.actuati'on prior to reactor'essel upper head void formatiiori in event of RCS diepressuriiation caused Iby a.steam line break.

These

are, indicated values that include. al'll'owances fair uncertainty.

The operator may manually bypass this tirip when pi.essurizer pressure is below 400 psia;'This

.bypass is automatically remi)ved when the pressurizer.'pressure increases to 500 psia.'

iLLrr,i -~gh The Containmept Pressure High trip provides assurance that a reactor trip i's initiated iii the evi nt of containment building. j)ressurization clue to a

pipe break inside thee icontainment bui1lding.

The setpoint for this tri~i is identical. to 'the safety injection setpoint.

G at Pr~s~r~-

ow

.The Steam Geneirator Pr'essure

'Low tt'ip provides protection in the event of an increa'se in heat removal'by the secondary'.

system arid subsequent cooldaw'ni of the reactor coolant.

The setpoint is sufficiently below, the,full 'load operating point so as not, to interfere with normal operation but still high.

enough to provide the'eqIuir'ed,protection in 'the. event df 'excesstyely high steam flow; This tt"ip's 'setpoint may be manually decreased as -steam generator pressure is reduced duririg plan't shutdowns, p'ro'vided'th'e margi'n 'between the steam-generator pressure and. this trip' setpoint is,maintained at. less tihan or equal to 200 psi; this sietpoint increases automatically" as steam-generator

'pressiire,-incr'eases anti.l the normal pressure trip setpoint is reached.

~vl ~.ow The Steam Generator Level Low tr'ip provides protect'i,on against a loss of feedwater.flow incident and assures that the",design-Ipressure"of

-the 'Re'actor Coolant Sy'tem-will not-be'xceeded.

due 'to a..decrease in. heat removal I'oy,thi.

,secondary system.

Yhi,s speci, fied-setpoint provides'allowance.that theire 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 auxiliary.feedwater is required to pr event degraded core cooling.

0 The Local Power Density

. High trip is,'provided. to,.prevent the.linear heat rate (kW/ft) in the limitin'g. fuel rod ini the core from exceeding the fuel:

design limit in the event of any design

bases, anticipated operational occurrence.

The local power-density is.caIlctilated in the reactor protective system,utiliiing the fol'iowing.information;.

PALO VERDE

. UNIT 3 B 2-4 Amendment No. 78

R SST S

Y 0 T SYSTEMS 0

T AK S

MIT NG I ION'O OP RATION'.

1. 1.1 The SHUTDOWN MARGIN shall be greater than or equal to that specified in the CORE OPERATING LIMITS REPORT.

MODES 3, 4* and 5* with the reactor trip breakers open**.

)ACT [gg:.

With the SHUTDOWN MARGIN less than that specified in the CORE OPERATING 'LIMITS

REPORT, immediately initiate and continue boration at greater than or equal to 26 gpm to reactor coolant system of a solution containing greater than or equal to 4000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEI L

CE RE UIREM NTS 4.1.1.1. 1'he SHUTDOWN MARGIN shall be determined to be greater than or equal to that specified in the CORE OPERATING LIMITS REPORT at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of at least the following factors:

1.

2.

3.

4.

5.

6.

Reactor Coolant System boron concentration, CEA position, Reactor Coolant System average temperature, Fuel: burnup based 'on gross thermal,,energy generation,,

'Xenon concentration, and Samarium concentration..

4. 1. 1. 1.2 The"overall core reactivity balance shall be compared to predicted values to demonstrate agreement within + 1.0X'elta k/k at least once per 31 Effective Full Power Days (EFPD).

This comparison shall consider at least those factors. stated. in Specification 4..1.1. 1. 1, above.

The predicted reactivity values shall be adjusted (normal. ized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD'fter each fuel loading.

.4. 1. 1. 1.3 With the reactor trip breakers open** and any CEA(s), fully or partially withdrawn, the SHUTDOWN MARGIN shall be verified within one hour after detection of the withdrawn CEA(s) and at least. once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the CEA(s) are withdrawn.

'See Special Test Exception 3. 10.9.

- The CEA drive system not capable of CEA withdrawal..

PALO VERDE UNIT 3 3/4 1-1 Amendment No. 8-,69,78

R ACT V TY C NTRO~SVS~TMS j' '&>>l i~AK i IMOB!'."

M T

-COND T ON FOlR OPERATION...

3;1.:1.2 The SHIUTDOWIN;-MARGINI shall be. greater, than, or equal to:that speci, fied

,in the 'CCIRE OF'ERIKTING LIMITS"REPORT;, -and; b.

c closed**

,~CT~O:

For T-~ less than or equal to 500'F, K., shall be less than 0.99.

Reactor criticality:shall not.be achieved. with shutdown group,CEA movement.

MODES I, 2',.'3*, 4*, and 5* with the reactor trip'reakersi URV a:

.With the SHIUTDOWN MARIGINI less than. that specified in the CORE OPERATING LIMITS,REPORT, immediately initiate and continue boration at -greater than or equal to 26 gpm to-the reactor coolant system of a solution conta,ining greater than or equial to-4000 ppm Iboron or'quivalent iiintil the required SHUTDOWN.MARGI¹-i's restored',

land,.

b.,With 'T.~ 'less than or equ il to',501);F *at>d<<K-., jIrea'ter-"'than'or. <<equal to 0;9!g, immediately, vary,CEA 'pcIsitia'tns.'and/or initiate and. cont'inue borat,i'iin:.at-greater'tt~an or equal to i26., gpm. to thi'e!actor coolant,i system of a solution containing, greater, than, or equal to 4000,ppm boron or equiv'alent uritil'the requireid'K.., is restored'.

4.1. 1.2.1

'With. the reactor'rip breakers cto'sed~~,

the..SIHUTDOWN, MARGIN,shall be determined to. be. greater than: oIr:equal to that,,speicified in the COFIE OPERATING LIMITS REIPORT:

a.

-Within '1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ..after--detection, ofl. an inoperable CEA(s), and a4 leapt

~

once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereaIFter w'hile thie',CEA(s).is inoperable.

See Special Test IExceptions 3.10.1 arid 3.10:9;

- The CEA-drive system capa,ble of CEA withdrawal.,

PALO VERDE UNI1 3

'3/4 1-2 Ainendment No&-,48, 6~

N N

R TAB 3.3-3 Contin ed SAF TY F ATU S

ACTU T N SY H"

N TR MENTATION T~NN N TNT ONE (a)

In NODES 3-4, the value may be decreased manually, to a minimum of 100

psia, as pressurizer pressure is reduced, provided:

(i) the margin between the pressurizer pressure and this value is maintained. at.less than or -equal to 400 -psi:;. and-(ii),.when the.RCS cold leg temperature is greater -than'or-equal: to 485 degrees F, this value is maintained at least 140 psi greater than the saturation pressure corresponding to the RCS cold leg temperature.

The setpoint shal.l be increased automatically as"pressurizer pressure is increased until the trip setpoint is reached.

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

(b)

In NODES 3-4, the 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.

(c)

Four channels

provided, arranged in a selective two-out-of-four configuration (i.e., one-out-of-two taken twice).

(d)

The proper two-out-of-four combination; (e)

.Input to channels.

The provisions of Specification 3.0.4 are not applicable.

T T T

ACTION 12 -

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

ACTION 13 -

With the number of channels OPERABLE one less thanthe. Total Number of Channels, STARTUP 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

bypassed, the desirability of maintaining this channel in the bypassed condition. shall be reviewed in accordance with Specification 6.5. 1.6.g.

The channel shall be returned 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 1.

Steam Generator Pressure Steam Generator Pressure Low Low Steam Generator Level I-Low (ESF)

Steam Generator-Level 2-Low (ESF) 2.

Steam-Generator Level (Wide Range)

Steam Generator Level - Low (RPS)

Steam Generator Level I-Low (ESF)

Steam Generator Level 2-Low (ESF)

PALO VERDE UNIT 3 3/4 3-23 Amendment No. 67, 78

~I

'TABLE.3..3" 3 (Co'ntinued)

ENGINEERED SAFEl'Y FEATUIRES ACTUATION SYSTEM INSTRUMENTATION ACTION STATEMENTS ACTION. 14-With the number of channels OPERABLE one less than the Minimum Channels

OPERABLE, STARTUP and/Ior IPOWER OPERATION::may. continue.

provicled thee following conditions are-sat'isfied a.

Verify that one of the inoperab'le channel's has.been bypassed aind 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 bypa.sed/tripped channel

. hall also be placed iin the bypassed/tripped condition as 1listecl below:

Process Measurement Circuit

>> Functional Unit Bypassed/Tripped 1.

Steam Generator Pressure. -

, Steam Generator'ressure

- Low Low i Steam Generator Level 1 -

Low (ESF)

Steam Generator Level 2 -

Low (ESF) 2.

Steam Generator (Wide Range)

STARTUP and/or POWER of the next required STARTUP and/or POWER restored to OPERABLE satisfiecl.

Level - Low Steam Generator Level -

ILow (IRPS)

Steam Cwnerator Level 1 - Low (ESF)

Steam Generator L'evel 2

'Low (ESF)

OPERATION may continue -iuntil the performance CHANNEL FUNCTIONAL'EST.

Subsequent, OPERATION may contintte if onie channel i'

status and the provisions of ACTION 13 are ACTION 15-ACTION 16 ACTION 17-ACTION 18-With the number of OPIERABLE channels one less than the Tota'l Number of Channels, restorie the inoperable'hannel 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 folio'wing i6 hours.

With the nuimber of OPIERABLIE channels one less tha'n the Total Number of Channels, bie in at least HOT STANDBY with'in 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SIHUTDOWN within the following.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ;

however, one channel may be bypass'ed'fo'r up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for surveillance testing !provided the other channell is OPERABLE.

With the number of OPERABLE channels one le'ss, than the Miinimum Number. of.Channels restore"the inoperable"channel" to"OPERABLE status within 48 hciurs 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 SHUTMiiitN.within the-follow'ing -3O hours.

With the number of OPERABLE channels one 'less than the Mini'mum Number of Channels,,

operation may continuie for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

After 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months operation may continue'r'ovided at least 1 train of essential filtration 'is in operation, otherwise, be in HOT'TANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 'nd

'in COLD SHUTDOWN witlhin the

'ollowing 30'ours.,

PALO VERDE " UNIT,3 3/4 3-24

R F T TUR S ACTUAT ON Y T M

NS RU T T TAB E NOTATION (I)

In. MODES 3-4, the. value may be decreased.

manually, to a minimum: of 100

psia, as pressurizer pressure is reduced;, provided:

(a) the margin between the pressurizer pressure and this value is maintained at less than or equal to 400 psi; and (b) when the RCS cold leg temperature is greater than or equal to 485 degrees F, this value is maintained at least 140 psi greater than the saturation pressure corresponding to the RCS cold leg temperature.

The setpoint shall be increased automatically.

as pressurizer pressure is increased until the trip setpoint is reached.

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

(2)

X of the distance between steam generator upper and lower level narrow range instrument nozzles.

(3)

In MODES 3-4, value may be decreased manual-ly 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)

X of the distance between steam generator upper and lower level wide range instrument nozzles.

PALO VERDE UNIT 3 3/4 3-27 Amendment No.

78

il k

C 0 G

YST HS S

M -

OP T

NG IMIT CO D TION FOR'PERATION 3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:

a.

One OPERABLE high-pressure safety injection pump, b.

One OPERABLE low-pressure safety injection pump, and Co

~OT ON:

'a ~

An independent OPERABLE flow path capable of taking suction from the refueling water tank on a safety injection actuation signal and automatically transferring suction to the containment sump on a

recirculation actuation signal.

NODES I, 2, and 3*.

With one ECCS. subsystem inoperable,, restore the, inoperable. subsystem to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the-fol.lowing 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

The current-value of the usage factor for each affected injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

With pressurizer pressure greater than or equal to 1837 psia, or RCS cold leg temperature greater than or equal or 485 degrees F.

PALO VERDE UNIT 3 3/4 5-3 Amendment No.

78

EMERGENCY CORE COiOLI YSTEMS SURVEILLANCE RE UIREMENTS 4.5.2 Each ECCS subsy. tern shall be demonstirat'ed OPERABLE:

At least once,per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by ver'ifying that. tlhe following valves are in the indicated positions with the ~valves key-locked shut:

Valve Number Valve Fun'ine'alve Position 1.

SIA HV-604

-1.

.HOT LEG INJECTION 1.

SHUT 2.

SIC HV-321,2.

'HOT LEG INIJECTION 2.

SHU'T 3.

SIB HV-609 3.

HOT LEG. IN'JECTION 3.

SHUT 4.

SID HV-331 4.

HOT LEG INJECTION 4.

'SHUT b.

At least once per 31 days Iby:

1.

Verifying that each valve (manual, power-operated, or automatic!)

in the f'low path that is not locked, seal'ed, or'therwise secured in po. ition, is in its correct position, and 2.

Verifying that tlhe ECCS piping is full of.water by.venting the accessible discharge piping high points.

c.

By a visual inspection Which verifies that no loose, debris (rags, trash, clothing, etc.) is present in 'th6 clontainment which could be transported to tlhe coI>tainment sumP and cause restriction of the pump suctions during ILOCA conditions.

Thi's visI>al inspection shall be performed:

1.

For all aiccessibTe areas of the conItainment prior to establ i shiing CON'll'AIINMENT INTEGRITY, and 2.

At least once daiily of the affected areas within containme'nt

'by'ontainment entry and during the. fina) entry when CONTAINMENT INTEGRITY is established,.

.d.

.At least once per 18 months by:

PALO VERDE - UNIT 3 3/4 5-4 AMENDMENT Nov 51

00 N

ST M

3 4.5.

U SY T MS SH T OWN MI NG CO D T ON fOR OP RAT ON 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a.

An OPERABL'E high pressure safety inj'ection pump, and'.

An OPERABLE flow path capable of taking suction from the refueling water tank on a safety injection actuation signal and automatically

. transferring suction to the containment sump on a recirculation actuation signal.

MODES 3* AND 4.

~C~:

With no ECCS subsystem OPERABLE, restore at least one ECCS subsystem to OPERABLE status within I hour or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b.

In the event the ECCS is actuated and injects wat'er. into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days'escribing the circumstances of the actuation and'he total accumulated actuation cycles to date.

The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

URVEI A CE RE U REMENTS 4.5.3 The ECCS subsystem shall be demonstrated OPERABLE per the. applicable surveillance requirements of Specification 4.5.2.

With pressurizer pressure less than..1837 psia. and".RCS. cold. leg, temperature less than 485 degrees F.

PALO VERDE UNIT 3 3/4 5-7 Amendment No.78

41 4k

~

/

ASES 4.3.1 and 3 4.3.2 REAC OR P

OT T AT ON SYST M

NST UM NTA I

AN ENGINE RED SAFETY FEATURES The OPERABILITY'fthe reactor protective and Engineered Safety-Features Actuation Systems instrumentation and.bypasses ensures. that (1) the associated Engineered Safety Features Actuation action. and/or reactor.trip: wi:,ll '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 safety analyses.

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-A, "RPS/ESFAS Extended Test Interval Evaluation,"

and CEN-327-A, Supplement 1,

and calculation 13-JC-SB-200-Rev.

01.

Response

time testing of resistance temperature devices;-which are"a part of the reactor protective system, shall be performed by using in-situ loop current test techniques or another NRC approved method.

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

Administrative controls on changes and periodic checking, of addressable constant val'ues (see also Technical Specifications 3.3.1 and 6.8. 1) ensure that inadvertent misloading of addressabl'e"constants into the CPCs is unlikely.

The design of the Control Element Assembly Calculators (CEAC) provides reactor protection in the event one or both CEACs become. inoperable.

If one CEAC is in test or inoperable, verification of CEA 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 CPCs in conjunction, with plant.

Technical Specifications will use DNBR and LPD penalty factors and increased DNBR and LPD margin to restrict reactor:operation, to. a.power. level:,that.will'nsure safe operation of the plant. If the margins are not maintained, a reactor trip will occur.

The value of the DNBR in Specification 2.1 is. conservatively compensated for measurement uncertainties.

.Therefore, the actual RCS total flow rate determined by the reactor coolant pump differential pressure instrumentation or by calorimetric calculations does not, have..to. be conservatively compensated for measurement uncertainties.

PALO VERDE'- UNIT 3 B 3/4 3-1 AMENDMENT'NO. 48, 50

SS-R AC T

~g'-Q(gg~gQ:-SA~F'Q ~QU~R~C~~~~Y~

N (Continued)

The measurement.

o',.;response time..at thi! sPedi.fied, frequencies. provides

,assurance., that.the,.-p'rotective,-;and..ESF

'action..functiion associated"with:,'each channel is completed'ithin the. time limit assumed in the safety analysds; No,:credit was taken. iin,,the, analyses for those cha'nnels with response tiii!es~

indicated as not'pplicable.

The instrumentationi rr sponse ti!iies are made up

.of,the time-,to gener aite the trip, signal at the detector (senscir

response, t;ime) and, the ',time.for the signal-to interruj!t power to the CEA drive mechanism (sigrial or trip del'ay time).

, Response time may. be demonstrated by any ser'ie's"o'f sequential, overlapping, or total channe'il test measurements.,proiiided:.-that such tests'emonstrate, the total charnel response time-as defined.

Sensor response time verification may be demonstrated by either (I) in p1lace, onsite,:or offsite*

test measurements, or (2) utilizincj replacement sensdrs,with certified response times.

Durin'g.normal operation,,

the low pressurider,.pressure trip,setpoint'ay

'e manually decreased, to a ii>inimum value of.160. p'saba;'s"pressurizer--pressure-is reduced -during plant shutdowns, provided the -margin between, the;,pressurize'r pressure and,this trip's setpoint is maintained.at less than'r i.qual to'400 psi; this. setpoint incrf!ases, automatically as pressuirizer pressure incre'ases

'ntil the trip setpoint is reaclhed.

.This se'tpoint must also be maintIin'ed

'at'east 140 psi greater, than the saturatiori pressure corresponding to the RCS cold leg -temperature whenever the RCS cold leg temperature is equal'o o'

greater than 485 degrees

.F.

This will ensure safety injection actuation prior, to reactor vessel upper head void. formation in event,.of-.RCS depressurization caused by a steam, line break.

These are indickted Valises'hat include allowances for uncertaiiiity.

The operator-ma'y iiianual.ly.bypass-.the low

'.pr'essurize'r,pressure trip when pressurizer. pressure is below 400 psia.

This

'bypass is automatirally removed when the pressurizer pressure increases to ~500 psia.

1I'al!K5MT¹N I~~O~O~~N~~ST i~U~NT~T~ON The OPERABILI1'Y iof.'.the rad'iation':monitoii'nig Channels"ensures:

that (I) the r'adiation levels, are continually measured in the areas served by'hie

'ndividual channels and (2) the alan> or autiomatic action: is initiated when the radiation level. trip siatpoint is exceeded.

JEQQKZ The OPERABILITY:iof the incore detectors~ with the-.specified minimum complement, of equipiment ensures

.that the-measuremi.nts obtained, from use

'of

'this.

system accurate1y represent the spatial neutron flux distribution of the'eactor core.

PALO VERDf UNIT 3' 3/4 3-2 Amendment Nb., 48-,89. 78

S S

U 0

p'C The OPERABILITY of the seismic. instrumentationensures that,.sufficient capability is available to promptly determine -the magnitude of a, seismi'c. event and evaluate the response of those features important'to safety.

This capability is required to permit comparison of the measured response

.to that used in the design basis for the.facil'ity'o determine if plant shutdown is required pursuant to Appendix A of 10 CFR Part,100.

The instrumentation is consistent with the recommendations of Regulatory Guide 1.12, "Instrumentation for Earthquakes,"

April 1974 as identified in the PVNGS FSAR.

The seismic instrumentation for the site is listed in Table 3.3-7'.

T 0

G CA STRUM NTAT N

The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data are available for estimating potential radiation doses to the public as a, result of routine or accidental release of radioactive materials to the atmosphere.

This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public. and is consistent with the recommendations of Regulatory Guide 1.23 "Onsite Meteorological Programs,"

February 1972.

Wind speeds less than 0.6 MPH cannot be measured by the meteorological:instrumentation.

T 0

S The OPERABILITY of the remote shutdown system ensures that sufficient capability is available to permit safe shutdown and maintenance of HOT STANDBY of the facil,ity from locations outside of the control room.

This capability is required in the event control room 'habitabil,i.ty, is lost and is consistent with General Design Criterion 19 of 10 CFR Part 50.

The parameters selected to be monitored ensure that (I) the condition of the reactor is known, (2) conditions in the RCS are

known, (3) the steam generators are available for 'residual heat removal, (4) a source of water is available for makeup to the

RCS, and (5) the charging system is avai.lable-to makeup water to the RCS.

The OPERABILITY of the remote, shutdown, system:.insures that a.,fire: wi:11.-

not preclude achieving safe shutdown.

The remote shutdown system instrumentation, control'nd power circuits, and disconnect swi.tches. necessary to eliminate effects of the fire and allow operation of instrumentation, control and power circuits required to achieve and maintain a safe shutdown condition are independent of areas where a fire could damage systems normally used to shutdown the reactor.

This capability is consistent with General Design Criterion 3 and Appendix R to 10 CFR 50.

The alternate disconnect methods or power or control circuits ensure that sufficient capability is available to permit shutdown and maintenance of cold shutdown of the facility by relying on additional'perator actions at local control stations rather than at the RSP.

PALO VERDE UNIT 3 B 3'/4 3-3 Amendment No. 48;~, 78

ST

- K33KIQ KliLIQKU5M The OPERABILITY-of the post-.accident nioni.tairihg. instrumentation, ensures that, sufficient-information, is available oA selected,plant:parameters to.

.monitor and assess'these'var'iables-fol1o'witig an:accident; This: capabil-ity=,is, consistent with ti'he recoirunendations of Regulatory, Guide 1.'97, "Instrumentation'for

'Light-'.Mater,-Cooled Nuclear P'lants

.to, Assess Plant, Condit~ions During and, Following an Accidetit,'" Diecember 1975 and'NURIEG 0578, "ilHI-2 Lessons Ldariied Task 'Force Status Report and Short-,Tenn Recomnendations;,"

The cont'ainment.'highi range area: monitors (RU-,148 R RU-'49) -and, the -main steamline radiatipn monitors (RU-139 AtlB and.le-140.ASB),are in Table.3.3463 The high range effluent, monitors and samplers (RU-;144'rid RU-146). are in the ODCH.

The containmerit hydrogen monitor's are in Specificatioti, 3/4.6.4. 1;-

The.

Post Accident Sampling System (RCS coolant) is in:lable 3..3-6.

'The Subcooleid hlargin Honi.tor

.(SHH):,. the I<eat

)unct;ion Thermocouple (HJTC),-

and the Ciore Exit Thermocouples (GET),comprise~ the Inadequate Core Cooling (ICC) instrumentation, required, by, Item II.I=.2 NUREG-()737, the frost THI-2 Action Plan.

The-function of the',ICC,.instrumentation's to enhancei the ability-of the plant, operator to diagnos~

t,he approach.-to existance of; and recovery from ICC.

Additionally they aid in tracking reactor coolant.

inventory.

These instruments are included: in the

,Iechn',ical.ipecificati ons.at

.the request of NRC Gieneric

-,I etter 83-37.,

These are not required by the accident. analysis, nor to bring the p'1lant to Cold Shutdown.

In the event more than four sensors'.in, ai Reactor Vessel Level channe'1 are inoperable, r'epairs may only be possibl'e during the next refueling outage.

This is because the sensors-are acces.ible only after the mi. sile shield

.and'eactor vessel head arie removed.

It is not feasib'le to repair a channel except during a refueling outage. when-the itiissile shield and reactor vessel head are,. removed to refuel

'the. core.

If btith chiannels are inope'rable; the channels shall be, restored to OPI=RABLE status iiii the nearest refueling, outage.

If only.one -chan'nel is.inoperable; i,t is iritended that,this channel be restored to, OPERABLE status in a refueling outage as soon-as reasonably possible.

~4..

OE -BLtILKl'j

'7 IL'I: ~NT N"

The OPERABILITY of, the loose-peart,detectiori,.instruliientation ensures; that sufficient. capability is availab'le to, detect 1'ouse metallic iparts in-the

primary system and; avoid or miitiigate damage to Itrimary system components.

The allowable,out-of-service tiimes and surveil'lance requirements are consistent with the recommendations.of.Regulatory Guide tl.133, ",Loose-Part Detection Program'or the Pririiary System of,Liglht-Water-.Cool'ed'Re'actors;"

-.Hay.1981.

. PALO VERDE

... UNIT 3

.B 3/4.'-4 Amendment No. 4'8-,34-,68,78

p

~

UMN T

N SES 33.

P ONI 0 RUNE TAT 0 I

The explosive gas. instrumentation is provided for, monitoring (and controlling) the concentrations of potentially.explosive gas mixtures in the GASEOUS RADWASTE SYSTEM; The OPERABILITY'nd use of'his instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

PALO VERDE UNIT 3 B 3/4 3-5 Amendment No. 78

4l 4i

~y

3/4. 5 EMERGENCY CORE COOLING SYSTEMS ECCS)

BASES 3/4..5.1 SAFETY INJECTION TANKS The OPERABILITY of each of the Safety Injection System (SIS}.safety injection tanks ensures that a sufficient volume of borated: water,.wil,l be immediately forced into the reactor core through each of the"cold legs, in the event the RCS pressure falls below the"pressure of the safety injection tanks.

This initial surge of water into the RCS provides the initial cooling mechanism during large RCS pipe ruptures.

The limits on safety injection tank volume, boron concentration, and pressure ensure that. the safety injection tanks will adequately perform their function in the event of a LOCA in MODE 1, 2, 3, or 4.

A minimum of 25K narrow range corresponding to 1790 cubic feet and a

maximum of 75K narrow range corresponding to 1927 cubic, feet of borated water are used in the safety analysis as the volume in the SITs.

To allow for instrument accuracy, 28K narrow range corresponding to 1802 cubic feet and 72K narrow range corresponding to 1914 cubic feet, are specified. in the Technical Specification.

A minimum of 593 psig and a maximum pressure of 632 psig are uses in the safety analysis, To allow. for instrument accuracy, 600 psig minimum and 625 psig maximum are specified in the Technical, Specification.

A boron. concentration of 2000 ppm minimum and 4400 ppm maximum are used in the safety analysis.

The Technical Specification lower 1'imit of 2300 ppm in the SIT assures that the backleakage from RCS will,not di lute the SITs below the 2000 ppm limit assumed in the safety anal'ysis prior to the time when drain-ing of the SIT is necessary.

The SIT isolation valves are not single failure proof;. therefore, whenever.

the valves are open power shall be removed from these, valves and the switch keylocked open.

These precautions ensure that the SITs are avai'labl'e during a Limiting Fault.

The SIT nitrogen vent valves are not single failure proof against depressurizing the SITs by spurious opening.

Therefore, power to the valves is removed while they are closed to ensure the safety analysis assumption of four pressurized SITs.

All of the SIT nitrogen vent valves are required to be operable so that, given a single failure, all four SITs may still, be vented'uring-post-LOCA long-term cooling.

Venting the SITs provides'or SIT depressurization capability which ensures the timely establishment of shutdown cooling entry conditions as assumed by the safety analysis for small break LOCAs.

The limits for operation with a safety injection tank inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with,fai,lure of an additional, safety injection tank which may result in unacceptable, peak cladding tempera-tures.

If a closed isolation valve cannot -be immediately-opened, the-ful.l capability of one safety.injection tank is;not avai:lable and prompt action is required-to place the reactor in a MODE.where this capability is not required.

For MODES 3 and 4 operation with pressurizer pressure less than 1837'sia the Technical Specifications require a.minimum of 57K wide range corresponding PALO VERDE " UNIT 3 B 3/4 5-1

SES~EN iJSI5 (C

to 1361'ubic feet and a maximum of 75X. narrow range corresponding-to 1927 cubic feet of borated water per tanlk, when three s'afety injection tanks are

'perable and a minimum of 36X wicle ',range corresponclirig to,908'ubic feet and a

maximum of 75X narrow range corresplonding to 1927'ubic 'feet. per tank when four,.safety injection. tanks are opelrable at, a minimum pressure, of 235 psig and a maximum pressure.;of 625,psig..

To allow for-instrumient.inaccuracy.,

-60X >ride range instrument corresponding tb= 1415 cubi@.feet;-

a'nd 72X. narr ow-,range instrument correspondir'ig to 1914 cub'ic 'feet',-when-three saf'ety, injection tanks are operable,.and'9X wide range instrument corresponding to 962 cubic feet, and 72X narrow range instrument correspor>ding to 1914 cubic feet, when four SITs are operableare specified in the Teclhnical Specifications..

To allbw for instrumerit,inaccuracy 2!54 psig is specified. in the Terhnical Specificatlions.

The instrumentation vs.

volume correlatidn.for'he 'SITs is, as, follows:

~1'me.

>trow ~an fide R~age

.962 ft;

<OX

'9X 1415 ft:'OX 60X'802 ft:-

28X

)8X 1914't'2X 83X'.5.~~~C~UBSYST~M!j The OPERABILITY of'wo separate and 'indepenlen't L'CCS'ubsystems with -the indicated RCS pressure greater than or equal to: 1837= psia;, or'-'w'ith the-indicated RCS cold leg temperature greater'than or equal to'85 degrees F

ensures that sufficient emergency core, cooling; capability'will *be:avai.lable in the event of.a LOCA, assuming the loss of on~ s'ub.'ystem t'hrough any single failure consideration.

These indicaited valises iriclude allowances

for, uncertainties.

Either subsystem oper'ating in conjunction with,the safety-injection tanks is capable 'of. upplying suflFicient core cooling to limit the peak cladding temperatures withiri acceptable'limits for all'ostulated break sizes ranging from.the.doilble-ended break o)F the. largest RCS cold 'leg p'ipe downward.

In additiion, each ECCS subs'ystem provi'des long-term, cori cooling capability. in the recircu'lation mode during'th'e acc'ident'ecovery.,period.

The. Mode 3 safety analysi.

credits one, HPSI puiap to provide negative

-reactivity insertion to -protect the co're and RCS'following a steam line break when RCS cold leg temperature is 485 degrees F oi greater.

Requiring two operable ECCS subsystems in the situation will ensure one HPSI jump is avai,lable,assuming single failure of'lhe other HPSI pump.

With the RCS cold"leg temperature b~low, 485 degr'ees;:F;,.one-OPERABL'E ECCS.

subsystem is acceptable without single failure cr)nsider'atic~n on the:basis of the stable reactivit.y condition of the reactor"ahd the lim>ted.cor'e

-coo'ling,'equirements..

The tr'isodium plhosphate

.dodecahydrate (TSP) stored in di'ssol'ving

'baskets located 'in the containment basement is provided'to min'imize, the 'possibility of

'orrosion cracking of certain metal components during operation of the ECCS following a LOCA.

Tlhe,TSP provided this priotection by dissolving in the sump water,and causing its f'inal

.pH to be rai'sedl t6"gaea'ter-than or equal to~

7'>0.~'he surveillancie requirements provided'o ensure OPERABILITY of -each

~

component ensure that at a min'imum, the assumptions.

u. ed iri the. safety analyses are met and that subsystem-OPERABILITY 'is maintai'ned.

Surveillance re'quiite-ments for -throttl,e valve,posit'ion stops, and flow -balance testing provide PALO VERDE - UNIT 3 B 3/4. 5-2 Amendment Nio.78

1 ADMINISTRATI E CONTROLS TS PO

.~ (~ (~<K

'6.9. 1.9 Core operating limits shall be established and documented in the CORE OPERATING LIHITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

a.

Shutdown 'Margin Reactor Tr'ip Breakers'Open for Specification 3.1.1.1

,b.

Shutdown Hargin - Reactor Trip Breakers Closed for Specification 3.1'.1.2 c.

Moderator Temperature Coefficient BOL and EOL limits for Specification 3.1.1.3 d.

Boron Dilution Alarms for Specification 3. 1.2.7 e.

Movable Control Assemblies - CEA '.Position for Specification 3.1;3.1 f..

Regulating CEA Insertion Limits for Specification 3. 1.3.6 g.

Part Length CEA Insertion Limits for Specification 3. 1.3.7 h.

Linear Heat Rate for Specification 3.2. 1 i.

Azimuthal Power Tilt T for. Specification 3.2.3 j.

DNBR Margin for.Specification. 3.2.4 k.

Axial Shape Index for 'Specification 3.2.7 l.

Boron Concentration (Mode 6) for 'Specification 3.9. 1 6.9.1.10 shall be a ~

b.

C.

d.

The analytical methods used to determine. the..core,.operating limits those previously, reviewed and approved by the NRC in:

"CE Method for Control'lement Assembly Ejection Analysis,"

CENPD-0190-A, January 1976 (Methodology for Specification 3.1.3.6, Regulating CEA 'Insertion Limits).

"The ROCS and DIT Computer Codes for Nuclear-Design,"

CENPD-266-P-A, April 1983 [Methodology for Specifications

3. 1. 1.1, Shutdown Margin Reactor Trip Breakers Open; 3. 1.1.2, Shutdown Margin Reactor Trip Breakers Closed;

3. 1. 1.3, Moderator Temperature Coefficient BOL and EOL limits; 3. 1.3.6,.Regulating, CEA Insertion Limits and 3.9.1, Boron Concentration (Mode 6)].

"Safety, Evaluation Report related to the Final Design of the Standard Nuclear Steam Supply Reference Systems CESSAR System 80; Docket No.

STN 50-470,

'"NUREG-0852 (November 1981),

Supplements No.

1 (March 1983),

No.

2 (September 1983),

No.

3 (December 1987)

(Methodol'ogy. for. Specifications..3. 1. 1.2,.Shutdown Margin Reactor Trip Breakers Closed; 3.1. 1.3, Moderator Temperature Coefficient BOL and EOL limits; 3. 1.2.7, Boron Dilution Alarms; 3.1'.3. 1, Movable Control Assemblies CEA Position; 3.1.3.6, Regulating CEA Insertion Limits; 3.1.3.7, Part Length CEA Insertion Limits and 3.2.3 Azimuthal Power Tilt - T ).

"Modified Statistical Combination of Uncertainties,'"

CEN-356(V)-P-A Revision Ol-P-A, May. 1988 and "System 80 Inlet Flow Distribution,"

Supplement 1-P to Enclosure 1-P to LD-82-054, February 1993 (Methodology for Specification 3.2.4, DNBR Hargin and 3.2.7 Axial Shape Index).

PALO'ERDE - UNIT 3 6-20a Amendment No. 42-,4~, 78

IN STRAT V~CONTRO~

JIPESLI59JJ I '&E'er tc

~h e.

"Calculative Methods.for the CE Larg!e Break LOCA Evaluation Model for the Analysis of CE and if Design!ed NSSS,"

CENPD-132, Supplement 3-P-A, June ]9S5'Methodo1logy for Specification 3.2.1,.Linear Heat Rate).

'f.

'Ca'Iculatfve Methods, for the ~CE~Small Break LOCA Evaluation Mode'!I,"

CENPD-137-'P,

August, I!974 (Methodology--for Specfff'cation;2.1,'inear Heat Rate).

"Calculatfv!e Methods for the CE Smhll Break LOCA Evaluatf!on Hodel',"

CENPD-137-P, Supplement 1P, Ja'nuary 1977 (Hethodo'logy for Speciffcatf<)n 3.2.1, Linear Hhati RIrte').

Letter::

0. D. Pairr (hlRC) to F. N.'Stern (CE), dated June 13, 1975 (NRC Staff Revieit of the Combustion Engine!ering ECCS Evaluation Model).

NRC approval for:

&.9.1.10f.

i.

Lett!er:

.K. Kniel (NRC) to A. E. Scherer (CE), dated Septemb!er 27, 1977 (Evaluation of Topical Reports CENPD-133,

~lupplement 3-P alnd'ENPD-137 Supplement 1-P);

NRC approval for. 6.'9'.1. 10.g.

The core operating 1limfts sihall.lbe determined so that,. all applicable 1fmfts (e.g.,

fuel thermal-mechanical 1 fmftscord, thermal-hydraulic 'limits',

ECC'8 limits, nuclear limits such as shutdown-margfi!r; and trarrsi'ent and an'alysi's 1 imits) of the safety analysis are met.

The CORE OPERATING LIMITS REPORT, including any mid-"cyccle revisions or supplements

thereto, sha11 be provided u~'ro'n's'su'ance, for each reload cycle, to the NRC Document

!Control Desk with 'coffee to the Regional Administrator'nd Resident Inspector.

PALO VERGE - UNIl' 6-20b AMENDMENT NO. -48, 55

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